ML20012A344

Exam Rept 50-395/OL-89-03 on 891205.Exam Results:Both Reactor Operator & Senior Reactor Operator Passed.Answer Key Encl

ML20012A344
Person / Time
Site:	Summer
Issue date:	02/02/1990
From:	Ernstes M, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20012A342	List: ML20012A341 ML20012A344
References
50-395-OL-89-03, 50-395-OL-89-3, NUDOCS 9003090284
Download: ML20012A344 (94)

Text

'

[-[ ' ' '

.y ;

p1Ed UNITED STATES i

L g.3 [#[", f ;

,%,g'

• f

= NUCLEAR REGULATORY COMMISSION i?

RE GION il-

-j.

101 MARIETTA STREET,N.W,.

4-h;;

I r4 t,

ATLANTA, GEORGI A 30323 ' '

,?'

i y<

7 t

u l[

ENCLOSURE 1

' ; s.

a 3

i

. EXAMINATION REPORT 395/0L-89-03 n

y?.

>i~

1 Facility Licensee: South Carolina Electric and Gas Company.

P. O. Box 88 i

s

~

Jenkinsville. SC 29065 Facility Name: _

V. C. Sumer Nuclear

Station

- i Facility Docket No.:

50-395 Facility License No.: NPF-12 Examinations were administered at V.

C.

Summer Nuclear Station near I

Jenkinsv111e, South Carolina.

J Chief Examiner: Mk

<t. b7M

/bo9o 1

-Michael E,,Ernstes Date Signed Approved By:'-

/

/

MQ

' John F. Munro / Chief

[ rate Signed Operator Licensing Section 1 1

Division of Reactor Safety-

SUMMARY

ExaminationswerehdministeredonDecember5,1989.

7 Written examinations were administered to one Reactor Operator and one sSenior Reactor 0perator, both of whom' passed.

1 I

h30gDOCK 05000395 O2e4 900202 5'

s

.V L

i PDC 4

2 3-

c t

j,'

+. g:

r f

i nl REPORT DETAILS 1.

Facility Employees Attending Exit Meeting:

. K. W. Woodward. Manager, Nuclear Operations Education and Training V. J. Kelly. Supervisor, Nuclear Operator Training.

G. Taylor. Manaoer. Operations 1

S. Furstenberg. Associate Manager, Operations G. A. Lippard, Nuclear Traininc Instructor 2.

Examiners:

• M. Ernstes. NRC, Region II

• Chief Examiner 3.

Exit Meeting:

At - the conclusion of the site visit, the examiner met with representatives of the plant staff to discuss the results of the examinations.

The licensee did not identify as proprietary any material provided to or reviewed by the examiners.

l' l.

i k

J

'5Arb PDI l

9. < s g <o9tr j

1 fapah) 013t j

V C; SUMMER SENIOR REACTOR OPERATOR' a

L ANNUAL REQUALIFIC ATION EXAMINATION-PART "A" SCENARIO #SS18 (ANSWER KEY)

FACILITY:

V.C. Summer i

REACTOR TYPE: PWR l

P DATE ADMINISTERED: 12/5/89 CANDIDATE:

i,

+

SECTION:-

TOTAL CANDII) ATE 'S POINTS POINTS A. Plant Operations 11.00 l

1 iH

. Submitted By:

//k 6-//

\\'

Date

//7 ~T-ST Approved:

[

/ ' Suhervisor,S' ulator And.

Date Training D elopment i

i 1

s l

p ',

~t e-L

2. -

pr,.: 93., -

y.

.1987 - 1988 Annual Requalification Examination

.)

L'*

lPart A (Answer Key) i a

l LQUESTIONi SS18-002 TIME: 8 MIN POINT VALUE:1.5 Match the following critical safety functions with the applicable level i

" ~ f challenge (i.e. Solid, Dashed, Dotted, or Satisfied).

(1.5)-

I 1

o M

a..

Suberiticality~

b.

LCore Cooling.

i

.,o c..

Heat Sink -

l d..

Integrity j

e.-

. Containment -

- f.. ' Inventory.

' ANSWER:

o

. a. -

Satisfied b.

. Dotted:

(0.25 points each)

,j c.

Dotted -

d.

Satisfied.

.t e.

- Satisfied-

f..

DottedJ COMMENTS:

Requires use of EOP-12.0 and comparison to existing plant conditions. RM-G7 reads-a little over 1R/HR but RM G17B reads 5 mr/hr, RM-G6 reads 5 mr/hr, RM-017A _

i

[

~ reads 6 mr/hr, and RM-G14 reads 5 mr/hr 50 CNTMT should be evaluated as SAT.

l

[

L l

t u

-b 4

N

c 7: 7 1987 1988 Annual Requalification Examination Part A (Answer Key)

. QUESTION: SS18-006 TIME: 5 MIN POINT VALUE:1.0 The condenser steam dump cannot be used to rapidly cool the RCS at the presenttime because:

(1.0) 7-a.

The condenser is not available at the present time.

b.

The condenser steam dump valves have no w'ay of being ARMED..

c.

There is still a valid STMLN AP HI actuation signal present.

d.

The Main Steam Isolation signal cannot be reset at the present time.

1 ANSWER:

d. (1.0).

COMMENTS:

Requires referencing EOP-4.0 an'd comparing,to plant conditions.

DO NOT USE with SS18-011. Distractor C is incorrect because high AP does not cause Mn Stm Isolation.

1 1

l-L

l g

, 14 1987 - 1988 Annual Requalification' Examination Part A (Answer Key)

QUESTION: SS18-010 TIME: 4 MIN POINT VALUE:1.0

Which of the following statements is TRUE concerning the steamline isolation.

. The steamline isolation has:

(1.0) a; Isolated the steam break

b..

Completely isolated the ruptured S/G.

c.

Limited the magnitude of the RCS Cooldown d.

Prevented steam flow from the faulted S/G ANSWER:

c(1.0)

COMMENTS:

Requires analysing conditions to determine beneficial effects of steamline isolation.

t l

l l:

ll l

l

.)

a n,;

h i

j y.3 y:3:a q..g.

y 1987 1988 Annual Requalification Examination j

~

- Part A (Answer Key) j QUESTIONi SS18-015 TIME: 5 MIN POINT VALUE:1.0 bc What emergency plan classification should be declared for the

[

eventin progress? -

(1.0)-

s a.

Notification of Unusual Event i

b. -

Alert

c..

Site Area.

j d.

General.

-ANSWER:

~ b. (1,0) (or c.)

COMMENTS: -

Requires analysing plant conditions and applyin,g EPP-001. Reference might 1

. be made to a LOCA m excess of capacity of enarging pumps in which case c would be correct.

a ll.

7 L

+

l

-[.

4

[

4 1

4

~

.c,..,_

1987 1988 Annual Requalification Examination' 4

Part A (Answer Key) e QUESTION:' SS18 018 TIME: EMIN-

- POINT VALUE:1'.0

--Which of the following motor-driven EFW pump start signals was -

L" generated fist?c (1.0).

a. _ ~ SI signal:

b.-

S/G low low level.

c.

Manual actuation

d.. -Trip of all Main Feedwater pumps ANSWER:

r d.-(1.0) g 2'

. COMMENTS:

Requires analyzing plant conditions to determine which of the applicable -

start signals are present. Then elimination of SI signal because SI sequencer.would not start EFW for 20 seconds.

p t

.I E

l:

l I

I-(;

f l

Y 4-e

n.

^

3. + ;.. ;;, ;.

i E

71987f 1988 Annual Requalification Examination Part A (Answer Key) 1 QUESTION: SS18 020 TIME: 4 MIN POINT VALUE:1.0 i

Select th'e response of the Pressurizer program level during this transient,

- up to and immediately after the automatic reactor trip.

(1.0) a.

Program level decreased until reactor trip, then remained

. constant.

b.

Program level remained constant until reactor trip, then -

decreased to no load value.

c..

Program level increased until reactor trip, then remained

constant, d.

Program level increased until reactor trip, then decreased to no load value.

ANSWER:.

b.' (1,0)

COMMENTS:

Requires recall that Pzr program level is function ofTavg and then.

' determining response of Tavg during transient. Ensure Pressurizer -

Level Program Trend Recorder is not in service PRIOR to using this question.

t

(

. b. ; u S

y., ;; S.

j 1987 - 1988 Annual Requalification Examination-:

Part A (Answer Key).

-QUESTION: SS18 022 TIME: 6 M1N-POINT VALUE:1.5 -

For the current plant conditions, what EFW flow rates are most appropriate for each S/G? Match the S/G in column A with the appropriate flow rate from -

f column B. Some choices in Column B may be used more than once.

-(1.5) a A

B C

li S/O'A; a.

Ogpm.

2.

S/G B b.

Enough flow to maintain 38% to 50% NR level '

3.

S/G C' c.

, Enough flow to keep S/G U tubes covered y

L d.-

Atleast300 gpm 0

e.

Atleast 500 gpm

. ANSWER:

1, b (.5)

2. a (.5)

'3*['hp.q1ft i

y '-r49 A

(vek COMMENTS:- -

Do not use with SS18-008 or SS18-009. Requires determining condition of S/Gs and then applying appropriate EOP.

t ~~

[

p

. (

y

.j 1987.1988' Annual Requalification Examination l

,,f

Part A (Answer Key) x

\\

LQUESTION: SS18-027 TIME: 5 MIN.

POINT VALUEil.0 l

Which of the following diverse parameters provide the most assurance that j

?

= a successful steam line isolation has occurred:

(1.0) j s

a.

' Annunciator points XCP-629 pts i-1,1-2,1-3

[

b.:. Steam flow on all three S/Gs is zero

~

m c.

Stable steam pressure on S/Gs A & C L-d.

S/G B steam pressure is zero psig

/ ANSWER:

c. (1.0).

COMMENTS:

Requires analyzing given information to determine which provides l

most assurance.

[

F i

h w

e

".. ~.

s 8

k v

~

y j

4 J~

.. p 1987 - 1988 Annual Requalification Examination-Part A.(Answer Key)

QUESTION:' SS18 029 TlME: 4 MIN

' POINT VALUE:1.0 Which of the following statements is TRUE concerning RB Instrument Air system.

(1,0)

a. -

RB Instrument Air is lined up and running normally r

b.

RB Instrument Air will eventually swap over to Station -

Instrument Air automatically 1

c.

RB Instrument Air is already being supplied by Station -

. Instrument Air.

d.

' RB Instrument Air will eventually be manually swapped over to Station Instrument Air.

. ANSWER:-

d.

(1.0)

COMMENTS:

Requires analyzing RB Instrument Air System and applying existing conditions.-

l l

l'

/

7:

$ 3 L

x

a 3

.c q..;,.

.1987 - 1988 Annual Requalification Examination Part A (Answer Key)

~ _ QUESTION: SS18-030 TIME: 4 MIN

_ POINT VALUE:1.0 -

l The status of the Station Instrument Air compressors is:

(1.0) t

a.

'A'IA compressor tripped because of SI and 'B'IA compressor started on low air pressure b.

. A'IA compressor tripped because of SI and 'B'IA compressor started on SI sequence.

{:

c.'

'A'IA compressor was manually tripped and 'B'IA compressor started on SI sequence.

d.

'A'IA compressor was manually tripped and 'B'IA compressor started on low air pressure.

ANSWER:

a.

(1.0)

COMMENTS:-

Requires analyzing existing conditions and applying system response to

.- SI.

5 I

)

.i

• ghpyohrl l

jf

'jQ 11tif -

i

+

JG/fuklAya T,

} V.C. SUMMER SENIOR REACTOR OPERATOR -

' : ANNUAL REQUALIFICATION EXAMINATION.-

PART "A" SCENARIO #SS19

~

(ANSWER KEY) k v

FACILITY:

V.C. Summer j

REACTOR TYPE: PWR DATE ADMINISTERED:'12/5/89 CANDIDATE:

SECTION:-

TOTAL CANDIDATE 'S POINTS POINTS A, Plant Operations 10.00 n

l o

j

. t.

//2-NI Appro"f.

///' s6pervihulatorAnd Date Trainin velopment L.

%. ? ;.;

t.

)

.. +.

19'87 1988 Annual Requalification Examination i

J Part A (Answer Key)

QUESTION: SS19-001 TIME: 5 MIN POINT VALUE:1.0 1

All of the following annunciator points could alarm on leakage outside of the RB EXCEPT:

(1.0) a.

XCP-606 PT-3 2 "LD TRBL RB/INCORE SUMP LVL HI" b.

XCP 607 PT-3 2 "LD RB SMP LVL HI" c.

XCP 621 PT-2-6 "VLV STM LKOFF/ SAT ANNUN TRBL"

+

d.

XCP-606 PT-2-2 "RBCU 1A/2A DRN FLO HI" t

ANSWER.

d. (1.0)

COMMENTS:-

Requires referring to ARPs to determine local annun panels which are alarming and the reference to ARP for local panels to determine if any inputs are exclusively RB leakage " excessive moisture after S/D" cause for d. eliminated by time at power in briefing. Do not use with SS19-002, i

s i

i f

l.

1987 1988 Annual Requalification Examination --

Part A (Answer Key) 1

QUESTION

SS19 004 TIME: 5 MIN POINT VALUE:1,0 Based on MCB indications, what is the location for the RCS leak?

(1.0) 7

.a.

SG "B" between the flow restrictor and the MSIV b.-

Letdown line, upstream of the regen HX c..

Charging line, upstream of the regen HX j

d.

Letdown line, downstream of the regen HX LANSWER:

b. (1,0)

COMMENTS:

Do not use with SS19 003. Letdown leak identified because the letdown.

flow is less than required by number of orifices. Leak upstream of regenerative heat exchanger identified because oflow charging temperature.

i y

s

}

j

7-i 1987 1988 Annual Requalification Examination Part A (Answer Key)

QUESTION: SS19 005 TIME: 5 MIN POINT VALUE:1,0 State whether the following parameters / valve pos?tions are greater than, less than, or the same as their expected positions.

(1.0) a.

LO PRESS LTDN Pressure PSIO PI 145 b

b.

LO PRESS LTDN PCV 145 c.

!!X divert control temperature TI 143 d.

CC to Ltda IIx TCV 144 ANSWER:

1 a.

Equsi to (0.25)

I b.

Less Than (0.25) c.

Less than (0.25) d.

Less than (0.25)

COMMENTS:

l Recpires comparisen of centrol b%rd values to norms L

l.

l, l

o 1987 1988 Annual Requalification Examination Part A (Answer Key)

QUESTION: SS19 006 TIME: EMIN POINT VALUE:1.0 Conduct a flow balance across the CVCS to estimate the RCS leak rate.

Choose the leak rate that is closest to actual calculated answer.

(1.0) a.

18 gpm' b.

42 gpm i

c.

66 gpm l

d.

96 gpm ANSWER:-

[

b.

(1.0) i t

COMMENTS: -

[

Charging pump output FI 122A 85 i2.5 gpm

= 85 i2.5 Sealinjection (FI 130A,127A,124A) 3 x 8 i.5 gpm

= 24 i 1.5

. LD HX outlet flow FI 150 57.5 i 2.5 gpm

= 57.5 i 2.5 RCP A sealleakoiY FR-154A 3 i.1 gpm

= 9 i.?. gpm l

- RCP B set Aleakoff FR 154A 3 i.1 gpm

[

/

RCP O sealleakofiFR 154A S i.2 gpm

= 42.5 i 6.5 gpm.

t COMMENTG:

Seal 14ection doesn't go thru FE 122 i

- e..- neg ects sealinject!or, i

.,,my

[I t

~.

l i

1987 1988 Annual Requalification Examination Part A (Answer Key) i QUESTION: SS19 007 TIME: 5 MIN POINT VALUE:1.0

- The dropped rod recovery was progressing normally when the annunciator XCP-620 pt 51 alarmed. Which of the following statements correctly

]

describes the cause and consequences of this alarm?

(1.0) i a.

The regulation failure on the opposite power cabinet will not i

stop rod J 13's motion.

b.

The regulation failure on Group 1 rods' power cabinet will stop rod J 13's motion.

l c.

The slave cycler failure in the logic cabinet will not stop rod J 13's motion.

~

d.

The slave cycler failure in the logic cabinet will stop I

rod J 13's motion.

ANSWER:

a. (1.0)

COMMENTS:

Reguires d2 termination ofintent of ARP, which only addresses logic cabinet failures.

i L

1 b

l l

l 4

}

l l

i r

i 1987 1988 Annual Requalification Examination Part A (Answer Key)

QUESTION: SS19 008 TIME: 5 MIN POINT VALUE:1.0 Is the reading on N-42 consistent with the current asymmetric rod i

position? Why or why not?

(1.0) a.

Yes itis. Rod J 13 has depressed the flux near N 42, b.

Yes it is. The radial Xenon oscillation caused by rod J 13 has depressed the flux near N 42.

c.

No it is not. Rod J 13 would depress N-44 more than N 42.

d.

No it is not. The central position of rod J 13 should depress flux across the whole core.

ANSWER:

c. (1.0) i COMMENTS:

a.

J 13 is closer to N 44

b. Xenon osciliat on hasn't had time to build in i

d. J-13 is a periphal rod Requires use of core map on MIDS panel to determine location of J 13 relative to NIs.

i t

/

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

I 1987 1988 Annual Requalification Examination Part A (Answer Key)

QUESTION: SS19 009 TIME: EMIN POINT VALUE:1,0 When rod J13 is realigned, the NROATC accidentally presses the rod control system START UP RESET pushbutton instead of the ALARM RESET pushbutton. Which annunciator would ni be expected as a resultof this mistake?

YW1 (1.0) a.

CMPTR ROD DEV/ SEQ NIS PR TILTS, XCP 620 pt 2-5 b.

CRB INSERT LMT LO LO, XCP 621 pt 1-1 c.

CRB INSERT LMT LO, XCP 621 pt 12

d..

ROD CONTROL URGENT FAILURE, XCP 620 pt 15 ANSWER:

d. (1.0)

COMMENTS:

Requires use of ARPs and system knowledge to determine effects of resetting group demand counters to zero at power.

p j

i 1987 1988 Annual Requalification Examination Part A (Answer Key)

QUESTION: SS19-013 TIME: 4 MIN POINT VALUE:1.0 If the bus IDA XFER INIT switch were to be placed in the E N position, the following would result:

(1,0) a.

Bus IDA would become inoperable due to transferring to Emergency power supply.

b.

Bus IDA would remain operable due to remaining on the Normal power supply, c.

Bus IDA would remain operable due to transferring to Emergency power supply.

d.

Bus IDA would become inoperable due to remaining on the Normal power supply.

ANSWER:

b. (1.0)

COMMENTS:

Requires evaluating Tech Suec caerahility based on conditions that exist when Tranafer Siitch [s manipulated.

l l

l i

1 l

1987 1988 Annual Requalification Examination Part A (Answer Key)

QUESTION: SS19-014 TIME: 5 MIN POINT VALUE:1.0 Given that the location of the leakage is currently unknown and has existed for the last hour, determine the Emergency Plan Classification of the current event.

(1.0) a.

NONE b.

NUE c.

Alert d.

Site Area ANSWER:

l

a. (1.0)

COMMENTS:

~

Anumed RC3 leskage p

I l

l l

77'-

y w-m w-v--e-n w

w-

=

,.ee

-r,--

.j p

1 1987 1988 Annual Requalification Examination l

Part A (Answer Key) j QUESTION: SS19 017 TIME: 5 MIN POINT VALUE:1.0 If Nuclear Instrumentation power range channel N41 where to instantaneously fall high, the resultant reactor trip would have l

occurred from:

.a.

Power Range High Flux, High Setpoint

[

l b.

S/G low. low level

c. -

Power Range High Flux Rate l

i d.

Power Range High Flux, Low Setpoint l

ANSWER:

l c.(1.0) f COMMENTS:

i Requires knowledge that negative rate bistable tripped on N42 and will meet required coincidence when positive rate bistable tr ps on N41. Do not use with SS19 021.

[

{

.2.

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

r r

V.C. SUMMER SENIOR REACTOR OPERATOR ANNUAL REQUALIFICATION EXAMINATION PART "B" WRITTEN (ANSWER KEY)

FACILITY:

V.C. Summer REACTOR TYPE: PWR DATE ADMINISTERED: 12/5/89 CANDIDATE:

i SECTION:

TOTAL CANDIDATE'S i

POINTS POINTS i

B. Limits'and Controls 23.00

's

/M -T-D Submitted By:

8 Date Approved:

/ /2 -r-87, s

[f[/

1. Sdpervisor,Sim lator And Date l

Training Dev pment

/

V LE

~ ~ -

~ 'o j

1987 1988 Annual Requalification Examination Part B (Answer Key)

QUESTION: OR 0228 TIME: 3 MIN POINT VALUE:1.0 Which one of the following reactivity coemeients is the most useful for determining the reactivity change that is required to produce a given change in Tavg when the reactor is operating at a constant power?

(1,0) t a.

Isothermal Temperature Defeat (Curve Book Figure 11-8.1) i b.

Isothermal Temperature Coemeient (Curve Book Figure 113.1, dashed line) c.

Moderator Temperature Coemeient (Curve Book Figure 113.1, solid line) d.

Total Power Defect (Curve Book Figure 112,1) j ANSWER:

c.

(1.0) i COMMENTS:

e, b Isothermal coemeient used for zero power,

d. NA at constant power i

f y

4 9

O s

1987 1988 Annual Requalification Examination Part B (Answer Key)

QUESTION: OR-0282 TIME: EMIN POINT VALUE:la At EOL, the reactor has failed to trip when required. The control room operators have taken actions as per the appropriate procedure (s) and obtained the required plant / system / component responses, except that the reactor is still not tripped, and emergency boration cannot be initiated because of blockage in the boration flow paths. All PR channels indicate 4%, and the startup rate is zero on both IR channels.

Which of the following describes the correct operator actions under these conditions '

AND the primary reason for taking those actions?

l 1

. a.

Return to the procedure and step in effect. Power is less than 5%, and the IR startup rate is sero, b.

Allow the RCS to heat up while continuing efforts to establish emergency boration. The bestup will j

insert negative reactivity.

Go to EOP.13.1. This is required by the suberiticality status c.

tree based on current reactor conditions.

I d.

Maintain RCS temperr.tures stable while continuing efiorts to establish einergency boration. Stable temperatures preclude reactivity insertion by cooldown.

ANSWER:

b. (1.0)

COMMENTS:

u a, c Solid path EOPs not excited until complete (step 12 RNO says i

L

" perform steps ofother EOPs", not "go to"

)

d. Not IAW EOP-13.0 (either H/U or C/D used, depending on MTC)

b. Also required determination that EOP-13.0 is the "appropria'te procedure"

t 1987 1988 Annual Requalification Examination a

Part B (Answer Key)

QUESTION: OR 0293 TIME: A MIN POINT VALgE:f A reactor startup is in progress. Power has been stabilised at 15/ib'to"* "

record critical data. Suddenly, all four Shutdown Bank B Group One rods drop.

i The proper response would be (1.0)

Pull Bank D rods to compensate and stabilise power, a.

i b.

Perform a recovery per the Abnormal Operating Procedures, Emergency borate to restore required shutdown margin.

c.

d.

Shutdown the reactor and make necessary repairs.

ANSWER:

i d.

(1.0)

COMMENTS:

a. is incorrect since T/S prohibit pulling Bank D rods

b. is incorrect since dropped group recovery is not discussed in AOP:

c. is incorrect since there has been no net loss of SDM i

9

)^-

3

.~.

.,..,_.,,,..._o.,

f U

}

1987 1988 Annual Requalification Examination Part B (Answer Key)

QUESTION: OR-0318 TIME: 1 MIN POINT VALUE:LQ The SOP.101 precaution II.4 concerns starting the second RHR pump during i

operations at half pipe. From the list below, select the correct basis for this precaution.

a.

Excessive flow through the reactor vessel.

b.

Starting pressure surge would disrupt low pressure letdown flow.

c.

Potential vortering due to decreased static suction head.

I i

d.

Excessive cooldown rate could add positive reactivity.

ANSWER:

i-i c.

(1,0) i COMMENTS:

Requires knowledge of bases for precautions, Short answer vers:.on of this question is OR-0062 i

l

)

0 i

1987 1988 Annual Requalincation Examination Part B (Answer Key)

QUESTION: OR 0322 TIME: EMIN POINT VALUE:LO The control room operators are responding to a SGTR with a loss of reactor coolant-subcooled recovery desired (EOP 4.2). Before they stop one charging pump (step 17) the following conditions exist:

One RCP running RCS hotleg temperature: 380'F RCS subcooling:48'F Two charging pumps running Containment pressure is 5 psig One RHR pump running PZR ievel 52%

After one charging pump is stopped, RCS subcooling is 35'F and PZR level is 45% and dropping.

The control room operators would:

I a.

Restart the second charging pump b.

Reinitiate SI l

c.

Establish normal charging d.

Depressurize the RCS to increase ECCS flow 1

ANSWER:

d.

(1.0)

COMMENTS:

Requires interpretation of meaning and intent of procedural step, recognition of adverse CNTMT

a. Only required if subcooling < 30'F

b. Not required. Also complicates plant control

c. Not appropriate with PZR level < [50%)

l i

a j

1987 1988 Annual Requalification Examination Part B (Answer Key) t QUESTION: OR 0355 TIME: 5 MIN POINT VALUE:LO i

The plant is operating at 100% power when offsite power is lost. The control room operators verify a reactor trip, turbine trip, and energization of the ac emergency buses. They ensure that the primary plant stabilizes at no load l

conditions and then check the secondary system. Water levelin S/G B is at 39% in the narrow range, and water level in the other S/Gs is below the i

narrow range. Which of the following is most correct concerning subsequent actions that the operating crew should take?

The operators should attempt to a.

Restore all S/Gs to between 38 to 50% narrow range level to allow more even cooling of the RCS. Adequate heat sink is already ensured with one S/G in the narrow range.

I b.

Restore all S/Gs to at least 38% narrow range level to achieve adequate heat sink AND to allow more even cooling of the RCS.

t Keep only S/G B at 38 to 50% narrow range level to ensure adequate c.

heat sink. Restoring all S/G levels would overcool the RCS.

9 d.

Keep only S/G B at 38 to 50% narrow range level to ensure adequate heat sink. Even cooling of the RCS is not a consideration.

ANSWER:

a. (1,0)

COMMENTS:

b. Heat sink is achieved w/one S/G.

l

c. Total EFW flow rate, not distribution, controls RCS cooldown rate.

l

d. Considered in background for ES 0.1

r t

l 1987 1988 Annual Requalification Examination l

Part B (Answer Key) t QUESTION: OR 0358 TIME: 3 MIN POINT VALUE:1.0 A reactor trip has occurred. The reactor trip and bypass breakers are open and neutron flux is decreasing. Rod bottom lights are on except for rods D4 and J2. They indicate full out. Select the appropriate operator action with respect to core reactivity.

a.

Borate using normal boration for 17 minutes for each stuck rod.

b.

Borate using normal boration for 17 minutes for the second stuck rod.

c.

Boratu using emergency boration for 17 minutes for each stuck rod.

d.

Borate using emergency boration for 17 minutes fbr the second stuck rod.

ANSWER:

c.(1.0)

COMMENTS:

T.S. requires > 30 pm of > 7000 ppm boron and EOP-1,1 requires borating for each stuck rod. This is an area not covered specifically in the AOP and therefore falls under objective 9. The student must go to EOP 1.1 not 1.0 to find actions for all rod bottom lights not lit.

P

[..

l i

1987 1988 Annual Requalification Examination l

Part B (Answer Key) i QUESTION: OR 0360 TIME: E MIN POINT VALUE:la Following a steamline break, the STA is manually monitoring the critical safety functions. At 14:00 hrs, he begins checking the Integrity status tree. RCS

)

cold leg temperature trending indicates the following:

j RCS Cold Leg Time

'i 12:00 hrs 13:00 hrs 14:00 hrs e

e e

' Loop A 410'F 320'F 230'F Loop B 415'F 320*F 22S'F Loop B 410'F 315'F 220F

?

Under these conditions, which of the following should the STA perform next?

a.

Check RCS pressure less than cold overpressure limit (425 psig).

s b.

Check the RCS pressure cold leg temperature points on the -

limit A curve.

c.

Recommend the CRS go to EOP-16.0, Response to Imminent Pressurized Thermal Shock Condition.

d.

Consider the CSF satisfied; go to the next CSFST, ANSWER:

a.(1.0)'.

COMMENTS:

a. Required by yes/no path for Integrity Tree,

b. Not required, CDR < 100 F/60 min.

c. Not required: No solid path on Integrity

d. Premature: required to check pressure first because < 350'F

1987 1988 Annual Requalification Examination Part B (Answer Key)

QUESTION: OR-0368 TIME: 6 MIN POINT VALUE:1.0 While the plant is operating in mode 1, a "PZR SAFETY VLV LINE TEMP

{

HI" annunciator is received. Leakage through one of the pressurizer safety l

valves is suspected. Pressurizer pressure and level are being maintained within normal operating bands by automatic control. Over the next hour, PRT levelincreases from 70% to 75%. Assuming identified leakage I

to the containment atmosphere is known to be 1.2 gpm the CRS should recommend:

i q

a.

Making preparations to be in Hot Standby within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

1 b.

Making preparation to be in Hot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Declaring the safety valve Inoperable and proceeding to I

i c.

Cold Shutdown.

d.

Cuntinued monitoring of temperatures and continued operation.

ANSWER:

I d.

(1.0)

COMMENTS:

Student must determine that leak rate is below T/S maximum and continued operation is allowed. A & B are incorrect interpretations of T/S requirements.

c is incorrect since the safety would not be declared inoperable until tested in mode 3. 70% = 7050 gal.75% = 7500 gal; 7500 7050 = 450 gal + 60 =

7.5 gpm 7.5 + 1.2 = 8.7 < 10 gpm T/S limit =+ No action

L

{

1987 1988 Annual Requalification Examination Part B (Answer Key) yh4 QUESTION: OR-0370 TIME: 4 MIN POINT VALUE L.0_

l 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after the pitnt shutdown, the RCS is at 140'F when a loss of the 2 running g" alfove the centerline and the vessel head is installed Ho 4

vesse *24 l

much time does the operating crew have to restore RHR before the core begins to unco */er?

ANSWER:

75 i 5 minutes COMMENTS:

Requires the examinee to recognize he is in a loss of RHR (EOP 2.4) and i

application of Att. H. Tolerance is given to 1/2 the smallest division, 4

~

+.--,a r

w

~

,4

E, L

1987 1988 Annual Requalification Examination Part B (Answer Key) l QUESTION: OR 0371 TIME: EMIN POINT VALUE:1_&

The control room operators are performing EOP-4.2, SGTR with Loss of Reactor Coolant Subcooled Recovery Desired. Offsite poweris not avallable. SI has been terminated and normal charging established per Step 18 but subsequently, the operators have determined that limited SI flow must be re-initiated to maintain adequate RCSinventory.

What is the proper way to re initiate SI flow?

a.

Cycle the reactor trip breakers to break the seel-in on the SI recet. This permits another auta SIif needed.

h.

Manuallyinitiate SI.

c.

Manually start the second enurging pump.

d.

Reopen HI HEAD to COLD LEG INJ valves MVG 8801 A & B.

ANSWER:

I d.

(1.0)

COMMENTS:

a,b results in undesired CNTMT ISOL, which complicates plant control.

c. Minimal effect w/ normal charging

d. Per step 22 alternate action, requires interpretation that" inability to maintain PZR level"is related to RCS inventory maintenance.

i j

1987 - 1988 Annual Requalification Examination Part B (Answer Key)

QUESTION: OR 0375 TIME: 5 MIN POINT val.UE:1.0 Offsite power is lost without SI actuation. The control room operators verify a reactor trip and a turbine trip. They determine that the D/Gs have re energized IDA and IDB. All appropriate loads have sequenced on. While ensuring that the RCS stabilizes at no load conditions, an operator observes that PZR pressure is 1980 psig and slowly decreasing. He checks the PZR PORVs and spray valves and notes that all are closed. PZR levelis stable at 25%

What corrective action,1f any, should be taken?

c.

No operator action s necessary. Pressure will stabilice at j

approximately 1850 psig.

i b.'

Manually actuate SI and return to EOP 1.0, Reacter Trip or Safety Injection Actuation, c.

Initiate a rapid depressurization to minimize RCS inventory l

loss through the RCP seals.

d.

Reset NON ESF LCKOUTS, then energize PZR Backup Heaters as necessary.

ANSWER:

d.

(1.0)

COMMENTS:

a. No indication of this Pressure is already below normal post trip level.

b. Not req'd. No indication ofloss of coolant

c. Only appropriate for loss of all AC in this case charging and/or CCW cools seals.

DO NOT USE WITH OR 0374

r.,

1987 1988 Annual Requalification Examination Part B (Answer Key)

QUESTION: OR 0376 TIME: EMIN POINT VALUE:LO The plant is operating at 100% when a fire in APN-5901 results in an overcurrent trip of the associated inverter (XIT-5901) output breaker.

l l

The plant does not trip.

l After the fire is out, which of the following actions should be taken?

a.

Re energize the bus immediately by reshutting the inverter l

output breaker.

W b.

Trip the plant and be h at least hot shutdown within C hours.

c.

Re-energize the bus immediately by shutting the alterna te

[

power supply breaker.

d.

Correct the problem and re-energize the bus withh4 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, f

ANSWER:

d.

(1.0)

COMMENTS:

a,c Re-energizing a faulted bus wouldjust cause further damage.

b. Not required, only one train ofinstrumentation affected.

d. Reg'd by T/S 3.8.3.1 action b.

Requires synthesis of knowledge of electrical grounds and T/S requirements, i

O O

1987 1988 Annual Requalification Examination Part B (Answer Key) l i

)

QUESTION: OR 0379 TIME: 4 MIN POINT VALUE:LQ During a normal surveillance, the following RWST parameters are determined e

Volume 458,360 gal i

e Temperature - 65'F e

Boron concentration. 2540 ppm Corrective action must be taken because:

i The basis for time to post LOCA switchover to hot leg recire is

[

a.

2500 ppm b.

Boron may precipitate out of solution at this temperature and 1

concentration.

i c.

Large volumes of water at this temperature would cause excessive thermal stress during injection.

I d.

This volume of water is insufficient for high and low head SI flow and RB Spray actuation.

ANSWER:

a.

(1.0)

COMMENTS:

This basis is not applicable to the T/S on the RWST as a borated water source but is applicable as a part of the ECCS.

o r

1987 1988 Annual Requalification Examination Part B (Answer Key) i QUESTION: OR 0383 TIME: & MIN POINT VALUE:la During the response to a main steamline break, a solid condition on the

{

Integrity status tree and a dashed condition (due to high containment pressure) on the Containment status tree are received. The control room operators perform EOP.16.0, Response to Imminent Pressurized Thermal Shock Condition, and establish an RCS temperature soak using the CVCS i

to maintain PZR level and using normal spray to maintain pressure.

During the soak, they enter EOP 17.0, Response to High Containment f

Pressure, and check the positioning of containment isolation phase "A" valves. Among the valves open is the letdown system containment Isolation valve, PVG-8152.

(1.0)

What should the CRS direct the RO ATC to do with PVG-8162, and why?

a.

Leave the valve cpen, because the letdown oriflees are the primary overpressure protection for the RCS.

b.

Leave the valve open, because the letdown isolation valves were deliberately opened to help control RCS pressure.

Close the valve: exiting EOP.16.0 indicates RCS integrity i

c.

is no longer a major concern.

i d.

Close the valve: this is necessary to preserve containment integrity.

ANSWER:

b.

(1.0)

COMMENTS:

Requires synthesis of knowledge related to RCS & RB integrity with several procedures.

a. Primary overpressure protection is PZR Safeties (PORV's backup).

b. Per FRG Z.1 step 1 background and E0P-16.0 step 20 l

c. Integri,ty is a, major concern during soak also,17.0 is not really " exited" 16.0 isin conjunction

d. Not required for CNTMT integrity: not an uncontrolled leak path I

during MSLB (CVCS intact).

Y

1987 1988 Annual Requalification Examination Part B (Answer Key) i i

QUESTION: OR-0384 TIME: 3 MIN POINT VALUE:1.0 The plant has tripped and safety injected, and the control room operators i

are responding to containment high pressure. During the response, j

utility management asks the ED to assess containment conditions. The i

ED reports that,"The potential fer flammable hydrogen concentrations

)

in containment la minimal....".

1 This statement would be accurate during a.

Inadequate core cooling, b.

A large-break LOCA in containment c.

A less ofreactor vesselintegrity d.

A main steamline break inside containment j

ANSWER:

d.

(1,0)

COMMENTS:

]

a, b, e contribute to hydrogen in CNTMT l

d. Would not be expected to contribute to hydrogen in CNTMT.

t l

4 D

0 1987 1988 Annual Requalification Examination Part B (Answer Key)

)

QUESTION: OR 0392 TIME: 6 MIN POINT VALUE:1.0 Component Cooling pump "A"is running, aligned to the active loop.

Component Cooling pump "C"is aligned to the inactive loop for performance ofits quarterly pump STP (122.002). Prior to starting pump "C", a safety injection occurs. (Lineup prior to SI: "A" pump running,"B" pump off with breaker racked up,"C" pump to "B"

{

train,oftwith breaker racked up). Which component cooling pumps will be running fellowing the safety injection, assuming no operator action?

(1.0)

A.

"A" pump enly B.

"B" pump only C.

"A" and "B" purnps only j

D.

"A", "B", and "C" pumps ANSWER:

A.

(1.0)

COMMENTS:

Interlock prevents starting either CCW pump on "B" Train therefore, only "A" pump will start.

i.

1987 1988 Annual Requalification Examination Part B (Answer Key)

QUESTION: OR-0395 TIME: 3 MIN POINT VALUE:1.0 Compared to the general SI termination criteria (i.e. following a small break LOCA), the termination criteria during a loss of emergency coolant recirculation are:

(1.0)

A.

Less restrictive for subcooling;less restrictire for inventory.

B.

Less restrictive for subcooling; more restrictuve for inventory.

C.

More restrictive for subcooling; less restrictive for inventory.

D.

More restrictive for subcoaling ; more restrictive for i'

inventory.

ANSWER:

C.

(1.0)

COMMENTS:

I l

1 I

r -..

1 i

!b 1987 1988 Annual Requalification Examination Part B (Answer Key)

QUESTION: OR 0405 TIME: 3 MIN POINT VALUE:la Which of the following should have already been accomplished prior to entry into EOP-4.47 (1.0) i 1

. I.

The ruptured S/G isolated II.

The RCS rapidly cooled down g,

III. The RCS depressurized l

1 i

A.

I. only B.

II. only C.

I and II enly i

i 1

'D.

I,II and III ANSWER:

C.

(1.0)

COMMENTS:

1 Requires evaluation of high level actions (big picture) of procedural transitions. Answer cannot be determined from EOP-4.4 entry " symptoms".

j i

l i

t_.

u

n,,

p

.)

7 hi[

1987 1988 Annual Requalification Examination y-Part B (Answer Key)

QUESTION: OR-0425 TIME: 5 MIN POINT VALUE:1.0 A LOCA is in progress with all RCPs secured, and the control room operators i.a are attempting to stabilize plant conditions. An operator who is monitoring l

- plant parameters observes the following:

RVLIS Narrow range 38%

e RVLIS upper head range 0%

Core exitTCs:

780'F e

RCS pressure 1920#

o L

Which of the following describes current core cooling conditions and operational requirements?

A.

Subcooled. Operator action is not required because core

.i t

4 coolingis satisfactory.

)

B.

Saturated. At their discrejon, the operators can take action to restore subcooled core cooling per EOP-14.2.

T C.

Degraded. Prompt action must be taken per EOP-14.1; or conditions could degrade to an inadequate core cooling condition.

D.

Inadequate. Prompt action must be taken per EOP-14.0, or core uncovery and fuel damage could occur.

ANSWER:

D.

COMMENTS:

Requires application of EOP-12.0 core cooling status tree criteria to a specific

- cass.

t t

i l\\

-ab'---

3

c..

e 1987 1988 Annual Requalification Examination-Part B (Answer Key)

QUESTION: OR-0429 TIME: 1 MIN POINT VALUE:LO The reactor is at power with all systems operating normally Control bank D is at 230 steps. Control rod H6 drops; the reactor does not trip.

r How'should the negative reactivity from the dropped rod be balanced out? (Assume a negative MTC).

l a.

Permit the plant to cooldown, adding positive reactivity.

l

b. -

Alternate dilute the boron concentration to add positive i

reactivity.

c.

Reduce power so that lowering the equilibrium Xenon concentration adds positive reactivity, i

?

d.

Reduce power so that lowering the power defect adds

~

positive reactivity.

i ANSWER:

I

d. -

COMMENTS:

a.' Required cooldown would be excessive.

j

b. Too slow to prevent excessive C/D l

c.: Xenon too slow; also, Xenon will initially increase on downpower.

l

' d. ' Requires interpretation of basic of AOP-403.6 step.

1 3

4

, a

s

-l m

1987 1988 Annual Requalification Examination Part B (Answer Key)

QUESTION: OR-0432 TIME: 3 MIN POINT VALUE:1.0 The reactor trips from 100% power with SI actuation. The cause is a L'

steamline break' downstream of the MSIVs. The control room operators

.take actions as per EOP-3.0 and obtain the required plant / system / component responses, except for the following: the MSIV associated with S/G A will not shut, and the maximum total EFW flow rate that can be achieved is 300 gpm.

.l S/G levels are as follows:

S/G A:

offscale low, narrow range S/G C:

45% narrow range S/G B:

offscale low, narrow range Which of the following correctly describes the impact of this configuration on

.i secondary heat removal capability?

a.

Secondary heat removal capability has been lost because EFW flow is inadequate.

b.

Secondary heat removal capability has been lost because S/G levels are inadequate.

1 c.

The loss of secondary heat removal capability is imminent, unless S/G A is isolated.

d.

Secondary heat removal capability is adequate at current S/G C levels.

. ANSWER:

d.

l COMMENTS:

Requires interpretation of basis of step to decide that S/G C level i's adequate for heat removal.

i l

p-

..t ;_

Qv a

1987 - 1988 Annual Requalification Examination Part B (Answer Key)

QUESTION: OR-0434 TIME: 3 MIN-POINT VALUE:1.0 The reactor trips from 100% power due to a spurious signal from the i

reactor protection system. The control room operators verify that the reactor is tripped and immediately note that the turbine is not tripped..

Because of a malfunction in the turbine trip system, none of the -

turbine stop, control, or combined intercept valves have closed, i

L Ifleft uncorrected, this malfunction is likely to result in The main turbine overspeeding, possible resulting in damage a.

to the main turbine rotor and shaft.

b.

A loss of condenser vacuum, resulting in the loss of condenser steam dumps, i

l c.

An increase in RCS pressure, possibly resulting in the PZR L

PORVs lifting.

d.

An uncontrolled cooldown of the RCS, resulting in reduced shutdown margin.

l ANSWER:

d.

COMMENTS:

a. Should be stopped by speed error on control & intercept vivs: also, generator has not tripped.

b. Not expected w/ cire, water pumps running.

c. Overcooling will decrease RCS pressure.

\\

N

' [

1,-

y&Ffor5Y

-'[

g v O..- :

i s k p

o. l s r ;

i Misba_ 07 5V

.V.C. SUMMER REACTOR OPERATOR l

iANNUAL REQUALIFICATION EXAMINATION

(

PART"A" SCENARIO #8818

-(ANSWER KEY)

FACILITY:

V.C. Summer REACTOR TYPE: PWR DATE ADMINISTERED: 12/5/89-CANDIDATE:

,s

(:

SECTION:

TOTAL CANDIDATE'S POINTS POINTS A. Plant Operations 11.00 Ee F

Submitted By:

/ /2 -f-8')

I Date Approved:/

///

//2 93Y s

[ '

' Supervi r, Simulator And Date C/

Tr

• ng Development N

A m

s,.

,s

^

=

g 1

(a;,'.;

+

- 1987 - 1988 Annual Requalification Examination:

' Part A (Answer Key)

. QUESTION: SS18-002 TIME:[8 MIN POINT VALUE:1.5 Match the following critical safety functions with the applicable level of challenge (i.e. Solid, Dashed, Dotted, or Satisfied).

'(1.5) a.-

Suberiticality b;

Core Cooling a

c.

Heat Sink d.

Integrity e.1 Containment f.

Inventory ANSWER:

a.

Satisfied

~

(0.25 points each) l b.

Dotted c.' -

Dotted

. d.

I Satisfied

e.

Satisfied

f. -

Dottedi COMMENTS:

Requires use of EOP-12.0 and comparison to existing plant conditions. RM G7 reads

' a little over 1R/HR but RM-G17B reads 5 mr/hr, RM-G6 reads 5 mr/hr, RM-G17A 4

reads 6 mr/hr, and RM-G14 reads 5 mr/hr 50 CNTMT should be evaluated as SAT.

?.

i L

9r -

m.s 1

y A

, e r

p e..;.,-

1987 1988 Annual Requalification Examination -

Part A (Answer Key)--

a QUESTION: SS18-004 TIME: 5 MIN POINT VALUE:1,0 Select the core exit temperature that the RCS would be rapidly ~

. cooled down to prior to depressurizing the RCS.

-(1.0) t a.~-

490'F -

b.'

505'F

. c.'

510'F.

4 d.

- 525'F J'

ANSWER:

a. (1.0) -

m COMMENTS:

-Requires applying EOP-4.0 step 17. Due to the ERG background document required knowledge that precise interpolation is not

. necessary, the temperature corresponding to the next higher -

L~

aressure in the tab le is not included. ADVERSE CONTAINMENT-

.DONDITIONS EXIST.

i I'

p l

L L.

[..

s l'

p-6 b

rr

7 k

1987 - 1988 Annual Requalification Examination -

Part A (Answer Key)

QUESTION: SS18-006 TIME: 5 MIN POINT VALUE:1.0 The condenser steam dump cannot be used to rapidly cool the RCS at the l

presenttime because:

(1,0) l a

a.-

The condenser is not available at the present time.

- b.

The condenser steam dump valves have no way of being ARMED..

j There is still a valid STMLN AP HI actuation signal present.

c.

i i

d.

The Main Steam Isolation signal cannot be reset at the i

present time.

i

~

-ANSWER:

)

d. (1,0) l L

i 1

COMMENTS-Requires referencing EOP-4.0 and comparing to plant conditions, i

DO NOT USE with SS18-011.- Distractor C is incorrect because high '

T AP does not cause Mn StmIsolation.

i 5

i

-l l l-L.

i -; :.

1987 1988 Annual Requalification Examination Part A (Answer Key)

F

QUESTION

SS18-017 TIME: 4 MIN-POINT VALUE:1.0

. Which of the following statements is TRUE concerning subsequent

. operation of the condensate pumps?

(1.0)

a..

Condensate pumps will trip b.

Condensate pumps are already tripped c.

Condensate pumps will go to minimum speed i

d.

Condensate pumps will stay running indefinitely ANSWER:

a. (1.0)

COMMENTS:

Rec uires analyzing conditions to determine that condensate pump wil:. soon trip on high DA level.

p l

l' o

l b

l j:

~

f

w g'

5 h.

> 1987' 1988 Annual Requalification Examination Part A (Answer Key)

QUESTION: SS18-019 TIME: 5 MIN-

' POINT VALUE:1.0 Select the correct choice that explains why the turbine driven EFW pump 1

t was started.

(1,0) a.

. Manual start

. b.

Low low levelin S/Gs c.

Trip of all MFW pumps j.

y d.

Undervoltage on 7200 VAC emergency buses 1DA & IDB 1

-i ANSWER:

i

a. (1.0) -

1 1

COMMENTS:

.i Requires analyzing plant conditions to determine which of the applicable u

start signals are present.

i i

i.

'I i

i

\\

)

r i

L U

p, T z,f t,

.1987 1988 Annual Requalification Examination Part A (Answer Key)

. QUESTION: SS18-020 TIME: 4 MIN POINT VALUE:1,0

. Select the response of the Pressurizer program level during this transient,

- up to and immediately after the automatic reactor trip.

(1.0) a.

Program level decreased until reactor trip, then remained constant.

s b.

Program level remained constant until reactor trip, then decreased to no load value.

c.

Program level increased until reactor trip, then remained constant.

d.

Program level increased until reactor trip, then decreased to no load value.

ANSWER:

b. (1.0)

COMMENTS:-

. Requires recall that Pzr program level is function of Tavg and then determining response of Tavg during transient. Ensure Pressurizer Level Program Trend Recorder is not in service PRIOR to using this -

question.

41

+

w-

+

7 1987 1988 Annual Requalification Examination i

o Part A (Answer Key)

. QUESTION: SS18-022 TIME: g MIN.

POINT VALUE:1.5 For the current plant conditions, what EFW flow rates are most appropriate for each S/G? Match the S/G in column A with the appropriate flow rate from column B. Some choices in Column B may be used more than once.

(1.5) o A

B~

1.

S/G A a.

O gpm 2.

S/G B'

b..

Enough flow to maintain 38% to 50% NR level 3.

S/G C c.

Enough flow to keep S/G U-tubes covered 5

d.

.Atleast 390 gpm e.

Atleast 500 gpm

. ANSWER:

l'. b (.5)

2. a (.5)

3. f(.5) -

(nkfU

. lk2 -r.-39 l

1. A*

. COMMENTS:

Do not use with SS18-008 or SS18-009. Requires determining condition of l S/Gs and then applying appropriate EOP.

l A

p --

.m

g,. '--

7.J9 1987 1988 Annual Requalification Examination 1

Part A (Answer Key) l t

QUESTION: SS18-025 TIME: 5 MIN

POINT VALUE:1.0 f

_Which'of the following is TRUE concerning the SAFETY INJECTION / PHASE A i

l ISOL(XCP-6104) status lights.

- (1.0) i a.

~Allindications are normal 1

b.

Two lights are dim that should be bright c.

One lightis dim that should be bright t

d.,

The status lights are NOT consistent with SI sequencer.

e

ANSWER:

a l(1.0) 4

. COMMENTS:

Requires applying status lights to determine if correct SI status Indication l-

. exists.

l.

L c

1, i

i i

4 m

1 1

T,: '..

1987' 1988 Annual Requalification Examination Part A (Answer Key)-

QUESTION: SS18 029 TIME: 4 M1N POINT VALUE:1.0

- Which'of the following statements is TRUE concerning RB Instrument Air system.

(1.0) a.-

RB Instrument Air is lined up and running normally b.

RB Instrument Air will eventually swap over to Station Instrument Air automatically c.

RB Instrument Air is already being supplied by Station H

Instrument Air.

d.-

RB Instrument Air will eventually be manually swapped over to Station Instrument Air.

o ANSWER:

L d.

(1.0) 1 COMMENTS:

Requires analyzing RB Instrument Air System and applying existing i

conditions.

Y l

I

~

~

~

1987 - 1988 Annual Requalification Examination Part A (Answer Key)

QUESTION: SS18 030 TIME: AMIN-POINT VALUE:1.0 The status of the Station Instrument Air compressors is:

,(1,0) n a.

'A'IA compressor tripped because of SI and 'B'IA compressor started on low air pressure b.;

'A'IA compressor tripped because of SI and 'B'IA compressor started on SI sequence.

c.

'A'IA compressor was manually tripped and 'B'IA compressor started on SI sequence, d.-

'A'IA compressor was manually tripped and 'B'IA compressor started on low air pressure.

I

. ' ANSWER:

I a.

(1.0)

COMMENTS:

.j Re i

l i i ti SI.qu res ana yz ng ex s ng conditions and applying system response to.

L l

i 1

I

j,( ] g* [Qyost 7

}w v.

3 3 774 j

t L

.. SUMMER REACTOR OPERATOR L

ANNUAL REQUALIFIC ATION: EXAMINATION-

^

-PART."A" SCENARIO #SS19 (ANSWER KEY):

p FACILITY:

V.C. Summer REACTOR TYPE: PWR DATE ADMINISTERED: 12/5/89 CANDIDATE:

i SECTION:

TOTAL CANDIDATE 'S POINTS POINTS-Af PlantOperations-10.00

/ /d'J-fY Submitted By:

Date

-- Approved:

/

//2-J

[

[ ~ 8 'buperviso,

imulator And Date Training Development

-- i

=

,. e 1987 1988 Annual Requalification Examination -

Part A (Answer Key)

. QUESTION: SS19-001_ TIME: 5 MIN POINT VA LUE:1.0 All of the following annunciator points could alarm on leakage outside L of the RB EXCEPT:

(1.0) a.

XCP-606 PT-3 2 "LD TRBL RB/INCORE SUMP LVL HI" b.

XCP 607 PT-3-2 "LD RB SMP LVL HI" c.

XCP-621 PT-2-6 "VLV STM LKOFF/ SAT ANNUN TRBL" d.

XCP-606 PT-2-2 "RBCU 1A/2A DRN FLO HI"

_ ANSWER:

d. (1.0)

COMME'NTS:

Requires referring to ARPs to determine local annun panels which are alarming and the reference to ARP for local panels to determine if any

' inputs are exclusively RB leakage " excessive moisture after S/D" cause for d. eliminated by time at power in briefing. Do not use with SS19-002.

L L

I' l

+a

.u.s a.--

a s-....

..s.-

s a

n.

4-r

.7

.: 2 1987 1988 Annual Requalification Examination Part A (Answer Key)

QUESTION: SS19 004 TIME: 5 MIN POINT VALUE:1.0 -

. Based on MCB indications, what is the location for the RCS leak?

(1.0) i a.-

SG "B" between the flow restrictor and the MSIV i

b.

Letdown line, upstream of the regen HX c.

Charging line, upstream of the regen HX

d..

Letdown line, downstream of the regen HX i

ANSWER:

b. (1.0).

3 COMMENTS:

Do not use with SS19-003. Letdown leak identified because the letdown :

flow is less than required by number of orifices. Leak upstream of regenerative heat exchanger identified because oflow charging temperature.

k i

a

e

.f.

c..,. :

- 1987 1988 Annual Requalification Examination 1

Part A (Answer Key)

QUESTION: SS19-005 TIME: 5 MIN -

- POINT VALUE:1.0 -

State whether the following parameters / valve positions are greater than, less than; or the same as their expected positions.

(1.0) a.

LO PRESS LTDN Pressure PSIG PI-145

' b.

LO PRESS LTDN PCV-145 c.

HX divert control temperature T1-143 d.

CC to Ltdn Hx TCV-144-ANSWER:.

a.-

. Equal to (0.25) b.

Less Than (0.25)

- c.

' Less than (0.25)

I l.-

. d.

Less than (0.25)

COMMENTS:

Requires comparison of control board values to norms 1

i L

l L

p 4

h ;

7,,

m g

1, n...

1987 - 1988 Annual Requalification Examination Part A (Answer Key)

~

1

! QUESTION: SS19-006 TIME: 5 MIN POINT VALUE:1.0 Conduct a flow balance across the CVCS to estimate the RCS leak rate.

l Choose the leak rate that is closest to actual calculated answer.

(1,0)

J

a..

18 gpm b.'

'42 gpm c.

66 gpm d.-

96 gpm -

ANSWER:~

-) b.

(1.0)

COMMENTS:-

Charging pump output FI-122A 85;i2.5 gpm

= 85 i2.5 Sealinjection (FI-130A,127A,124A) 3 x 8 i.5 gpm

= 24 i 1.5

~

' LD HX outlet flow FI 150 57.5 'i 2.5 gpm '

= -57.5 2.5 :

2 RCP A'sealleakofTFR-154A 3 i.1 gpm

= -9 i.3 gpm RCP B sealleakoff FR-154A 3 i.1 gpm RCP O sealleakoff FR-154A 3 i.1 gpm

= 42.5 -i 6.5 gpm

' COMMENTS:

Sealinlection doesn't go thru FE 122

a., neg.ects' seal injection i"

l

h

-,j.,.

1987 1988 Annual Requalification Examination Part A (Answer Key)

L

. QUESTION: SS19 007 TIME: 5 MIN POINT V ALUE:1.0 i"

The dropped rod recovery was progressing normally when the annunciator

^

XCP-620 pt 51 alarmed.- Which of the following statements correctly l describes the cause and consequences of this alarm?

(1.0) 1 a.

The regulation failure on the opposite power cabinet will not j

stop rod J-13's motion.

I b.

The regulation failure on Group 1 rods' power cabinet will stop rod J 13's motion.

c.

The slave cycler failure in the logic cabinet will not stop rod J-13's motion, d.

The slave cycler failure in the logic cabinet will stop rod J-13's motion.

ANSWER:

a. (1.0) t COMMENTS:

Requires determination ofintent of ARP, which only addresses logic cabinet failures.

+

b

g-

)

L

~ 1987_.1988 Annual Requalification Examination j

["

Part A (Answer Key) l

' QUESTION: SS19-008 5.'IME: 5 M1N POINT VALUE:1.0 Is the reading on N-42 consistent with the current asymmetric rod

position? Why or why not?

(1.0) a.

Yes itis. Rod J-13 has depressed the flux near N-42.

b.

Yes it is. The radial Xenon oscillation caused by rod J-13 has depressed the flux near N-42.

c.

No it is not. Rod J-13 would depress N-44 more than N-42.

[

d, No it is not. The central position of rod J-13 should depress flux across the whole core.

ANSWER:

c.(1.0)-

? COMMENTS:

a.

J-13 is closer to N-44

b.. Xenon oscillation hasn't had time to build in

d. J-13 is a p,eriphal rod Requires use ofcore map on MIDS panel to determine location ofJ-13 relative to NIs.

l t

4 I

l r

'E -

f, l

1 3,.

n*.+

l 1987 1988 Annual Requalification Examination' Part A (Answer Key);

i j

9-

. QUESTION:' SS19-010 TIME: 5 MIN POINT VALUE:1.0 1

. Assume that a quadrant power tilt ratio was calculated 30 minutes ago and was determined to be 1.019. Select the technical specification whose

' time limit would be exceeded first. Assume that no other actions were taken.

(1.0);

a.

3.1.3.1, Movable Control Assemblies Group Height b.

3.2.1, AxialFlux Difference (AFD) y

c. -

3.2.4, Quadrant Power Tilt Ratio d.

3.4.6.2, Operational Leakage t'

1 ANSWER:-

a. (1.0) -

C'OMMENTS:

- Requires applying T/S to current plant conditions.

1 I

i i

l l '-

{

i=

l

7 p

' 'E., g, ' -

1987' 1988 Annual Requalification Examination Part A (Answer Key)

. QUESTION: SS19-012 TIME: 5 MIN POINT VALUE:1.0

' Assume that N42 has been declared inoperable and is being removed from service per the applicable AOP. The instrument power fuses are mistakenly removed instead of the control power fuses for N-42. As a result of this error, the following will occur:

(1.0) a.-

The channel will indicate 0%, both on the drawer and at its remote indications, but the protective bistables will remain

, in their existing condition, b.

Only the N-42 drawer indications will be affected. All remote indications and output bistables will respond' normally to 0% power output.

c.

The channel will indicate 0% output, and the reactor trip ab Iigh flux high and low setpoints and high rate) bistables will go to the tripped condition. The P-8, P-9, and P-10 bistables will respond normally to a 0% power output.

d.'

The channel will indicate 0%, and all ofits bistables will go to the tripped condition.

ANSWER:

d. - (1,0)

COMMENTS:-

Ll

/

8

g q

(?

u.

., L 'c ' -., _.,

1987 1988 Annual Requalification Examination.

u Part A (Answer Key)

- QUESTION: SS19-013 TIME: 4 MIN POINT VALUE:1.0 If the bus IDA XFER INIT switch were to be placed in the E N position, the following would result:

(1.0) a.

Bus IDA would become inoperable due to transferring to Emergency power supply.

b.. - Bus IDA would remain operable due to remaining on the Normal power supply.

c.

Bus 1DA would remain operable due to transferring to Emergency power supply.

d.

Bus IDA would become inoperable due to remaining on

' the Normal power supply.

t ANSWER:

L

b. (1.0) l l'

i

?

COMMENTS:

Requires evaluating Tech Spec operability based on conditions that exist when Transfer Switch is manipulated.

l

- 2

i' --

1987 1988 Annual Requalification Examination Part A (Answer Key)

QUESTION: SS19-018 TIME: 3. MIN POINT V A LUE:1.0 Regarding ROD SPEED STEPS / MIN (SI-408), which is the most correct concerning the indicated rod speed?

(1.0).

3 a.

Consistent with existing plant conditions b

b.

Should be reading 48 steps per minute c.

Should be based on existing Tavg-Tref deviation d.

Should be reading 72 steps per minute i

ANSWER:

a. (1,0)

. COMMENTS:

. Rod control switch selected to shutdown Bank.A

(:

1

/ y". s m

r

, og... m..;

V.C.' SUMMER REACTOR OPERATOR f

. ANNUAL REQUALIFICATION EXAMINATION:

PART "B"~ WRITTEN (ANSWER KEY) t r

FACILITY:

V.C. Summer REACTOR TYPE: PWR-DATE ADMINISTERED: 12/5/89 CANDIDATE:

i

. SECTION:

TOTAL CANDIDATE 'S POINTS POINTS

.4

. B.' Limits and Controls 23.00 l

i 4

. Submitted By:

/

//c2 4-f/

Date

' Approved: ^

/ /7 - >-Q Supervisor, Si ulator And Date Training D opment l

)

L1987 - 1988 Annual Requalification Examination j

Part B (Answer Key)

QUESTION: OR-0211 TIME: 5 MIN POINT VALUE:1.0

- While the plant is operating at 100% power, the CCW header B radiation monito:'IRi-L2B falls low, and the "CC LOOP B RM.L2 B TRBL" annunciator alarms. Which of the following describes the restrictions,if any, on continued operation of the CCW system?

(1.0)

-j

a..

Operation can continue if, at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, grab.

samples are collected and analyzed for gross radioactivity

. (beta or gamma) at a limit of detection of at least 10MW* d

~

microcuries/ml.

b.-

Operation can continue for up to 30 days, provided that the surge tank liquid level is monitored for

~f

. unanticipated level changes and logged at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, c.

Operation can continue as long as the CCW header A radiation monitor RM L2A remains operable.

-l

.d.

There is no technical specification LCO associated with j

the CCW header radiation monitors; operation can H

proceed without restriction.

- ANSWER:

.d.

(1,0)

COMMENTS:

Requires checking Tech. Specs. for Rad monitors, Liquid effluent rad

- monitors, and RCS leakage detection systems. Some "no action" answers are required in exam bank to maintain "no action" as a credible distractor.

I,,

t

_ 1987 1988 Annual Requalification Examination Part B (Answer Key)

QUESTION: OR 0293 TIME: E MIN POINT VALUE:.LQ A reactor startup is in progress Power has been stabillied at 10Y)% to record critical data. Suddenly, all four Shutdown Bank B Group One rods drop.

The proper response would be (1,0) a.

Pull Bank D rods to compensate and stabillie power.

b.

Perform a recovery per the Abnormal Operating Procedures.

c.

Emergency borate to restore required shutdown margin.

d.

Shutdown the reactor and make necessary repairs.

ANSWER:

d.

(1.0)

COMMENTS:

a. is incorrect since T/S prohibit pulling Bank D rods b is incorrect since dropped group recovery is not discussed in AOPs

c. is incorrect since there has been no net loss of SDM

---

...--.-...,--.-..,.-.a.

I o

1..

i 1987 1988 Annual Requalification Examination i

Part B (Answer Key)

{

QUESTION: OR 0302 TIMEt ft MIN POINT VALUE:LQ The control room operators are responding to a SGTR with PZR pressure l

control available. To cool down the RCS and establish subcooled margin, l

the operators dump steam to the condenser using an intact S/O.

RCS cooldown using steam dumps is the preferred method compared to l

dumping steam through the PWR RELIEF of an intact S/G primarily because of which of the following?

(1.0) a.

It minimizes radiological releases and conserves CST inventory, f

b.

It minimizes primary to-secondary leakage and is easier to control.

l c.

It promotes S/G filling, preventing ruptured S/O dryout.

d, Steam dumps can be isolated more easily if they fall open than the PORV's can.

ANSWER:

a.

(1.0)

COMMENTS:

b. Not directly (pressure diff., not temp., determines leak rate) control similar

c. Ruptured S/O will fill, not dry out

d. Isolation for both is similar (manual valve)

WOO d. deleted, also true & important (12.2% vs 7.9% N)

l 1987 1988 Annual Requalification Examination Part B (Answer Key)

QUESTION: OR 0343 TIME: EMIN POINT VALUE:LO The reactor has tripped due to a loss of MFW, and the control room operators initiated RCS bleed and feed heat removal because of a loss of secondary heat i

sink (EOP 15.0). The operators are preparing to terminate bleed and feed because secondary heat sink has been restored and verified. The operators check if a charging pump can be stopped; they are unable to stop a t

charging pump because RCS subcooling is 25'F (Step 19).

i Although RCS subcooling is insumcient right now, subcooling will begin to increase as RCS pressure increases due to subsequent a.

Closing of the PZR PORVs.

b.

RCS heat up.

I c.

Establishment of normal charging flow.

d.

Closing of the PZR spray valves.

ANSWER:

a. (1.0)

COMMENTS:

b. Step 18 verifles that the RCS is not heating up

c. Reducing charging flow from cold leg injection to normal charging will decrease pressure

d. PZR spray valves are already closed by procedure.

I

+

--.m-s-.---*----

m

i 1987 1988 Annual Requalification Examination

[

Part B (Answer Key)

QUESTION: OR-0345 TlME: EMIN POINT VALUE:1A The reactor has tripped due to a loss of MFW, and the control room operators l

are responding to a loss of secondary heat sink. RCS bleed and feed is in progress, and the operators are preparing to restore EFW flow to S/G A (Step 16). The following conditions exist:

I Core exit'Its:

Approximately 610'F and stable Loop A hotleg temperature:

590'F and stable l

S/G A wide range level:

5%

l Under these conditions, which of the following EFW flow rate.is the most correct?

a.

Maximum rate to refill S/G A as quickly as possible.

b.

Maximum rate required to satisfy the heat sink status tree (390 gpm).

c.

Very slow rate to minimize thermal shock to S/O A.

d.

Maximum rate, consistent with NPSH available (CST height).

Slow enough to prevent runout of the EFW pump.

e.

ANSWER:

c. (1.0)

COMMENTS:

a,b,d Could result in S/G tube sheet failure due to thermal stress (21" plate,120'F EFW to > 550'F Primary coolant)

~.

QUESTION: OR 0346 TIME: 5. MIN POINT VALUE:LO The control room operators are responding to a loss of all ac power.

They begin a controlled depressurization of the intact S/Os to minimize RCSinventoryloss.

During the S/O depressurization, they monitor RCS TC primarily to ensure that the depressurization DOES NOT Disrupt the natural circulation occurring in the RCS loops.

a.

b.

Challenge the RCS integrity critical safety function.

c.

Create voiding in the reactor vessel upper head and loss of pressurizer level.

i d.

Overcool the RCS, thus permitting patential introduction I

of nitrogen from the accumulators into the RCS.

l j

ANSWER:

b. (1.0)

COMMENTS:

a. Continuous secondary depressurization should increase, not disrupt, NC.

c. Cooling the RCS should increase (or at least maintain) RCS subcooling:

voiding not expected w/o inventory loss. Voiding would increase PZR level,

d. Nitrogen injection is a function of RCS pressure: thisis limited by keeping S/G pressure > 250 psig.

4

+

1987 1988 Annual Requalification Examination Part B (Answer Key)

QUESTION: OR 0347 TIME: 1 MIN POINT VALUE:LO f

The control room operators are responding to a loss of all ac power (EOP 6.0).

they have verilled a reactor trip and a turbine trip and are in the process of l

isolating the RCS (Step 3). They verify that the PZR PORVs and letdown isolation valves are closed and then check the excess letdown isolation valves, both of which are open.

Ifleft uncorrected, this excess letdown isolation valve misalignment could l

create a leak path from RCS loop C.

To the PRT via the letdown line relief valve. RCS inventory loss a.

could increase, reducing the time to PRT overfill / rupture.

b.

To the PRT via the RCP seal return relief valve. RCS inventory loss could increase, reducing the time to core uncovery.

To the RCDT via the RCP #2 sealleakoffline. RCS inventory loss c.

could increase, possibly leading to core uncovery, d.

To the RCDT, increasing the RCS leakage rate until it is terminated by automatic isolation upon PZR low level.

ANSWER:

b. (1,0)

COMMENTS:

b. Requires use of flow diagrams to determine possible unintended flowpaths,

a. Excess letdown doesn't go to letdown line.

c. #2 sealleakofialso goes to PRT

d. No flowpath from excess letdown to RCDT,

i 1987 1988 Annual Requalification Examination Part B (Answer Key)

QUESTION: OR 0349 TIME: 5 MIN POINT VALUE:1.0 i

A loss of all ac power has occurred. The control room operators have

^ performed the immediate actions of EOP 6.0 and have attempted, without success, to restore power from the control room. As directed by EOP-6.0 Step 6, the operators place various control switches in the PULL-TO-LOCK position, including the charging pump switches.

The defeat of the charging pumps' automatic loading is designed to prevent bus overloading, to prevent pump damage when ac power is locally restored, and to prevent a.

The delivery of cold seal injection flow into the RCP no.1 seal chamber and shaft area.

b.

The unnecessary use of water that may be needed forlong term plant recovery.

c.

An uncontrolled overpressurization of the primary, and the i.

resulting increased loss of RCS inventory through the -

L RCP seals.

~

d.

'An uncontrolled cooldown of the primary due to the injection of cold water into the RCS, possibly leading to a reactor restart.

ANSWER:

a. (1,0)

I*

COMMENTS:

L-

b. Charging is the most necessary use of RWST

c. Not ex sected if charging is restored; RCS will be hot, near norma., no load Tavg =+ charging flow low, no PTS concerns.

d. RCS Tavg being determined by S/G safeties; little effect from charging.

P 1987 1988 Annual Requalification Examination Part B (Answer Key)

QUESTION: OR-0351 TIME: 5 MIN POINT VALUE:LO A loss of all AC power has occurred. The control room operators have performed the immediate operator actions of EOP-6.0 and have unsuccessfully attempted to restore power from the Control Room.

They discuss whether or not the SI signal that is present should be reset. RCS pressure is below the SI actuation setpoint, and RCS subcooling and PZR level are above the minimum values that permit recovery without SI.

Should the SI signal be reset? Why or why not?

a.

Yes. The signal is reset to permit loading of the ac emergency bus. The ac bus cannot be manually or automatically loaded with the signal present, j

i b.

Yes. The signal is reset to allow manual starting of ECCS equipment when poweris restored and to permit recovery per EOP.6.1.

L c.

No. The procedural caution concerning resetting SI does not apply for the described conditions. A valid signal should not be reset.

L l

d.

No Resetting SIis not necessary under these conditions because the signal seal in was broken when IDA and IDB where deenergized.

ANSWER:

b. (1.0)

COMMENTS:

a. The bus will automatically load in SI sequence of ESFLS: point is to defeat auto start.

c. Not true per ERG background for blackout.

d. Instrument buses remained powered by batteries thru inverters.

i lL

~*

I 1

1987 1988 Annual Requalification Examination Part B (Answer Key) l QUESTION: OR-0352 TIME: 3 MIN POINT VALUE:14 The plant is operating at 100% reactor powcr when offsite electrical power

{

is lost. The D/Gs start and energize the ac emergency buses, and all emergency equipment starts as required. SIis not required and does not actuate. Hewever, offsite power remains unavailable.

Which of the phrases listed below correctly completes the following statement?

Ten minutes after the event while offsite power is still unavailable, coolant flow is i

driven between the core and the S/Gs by n.

Injection flow from the ECCS pumps.

I b.

Forced circulation from the RCPs.

c.

Natural circulation i

d.

Reflux boiling.

l ANSWER:

I

c. (1,0)

I l

COMMENTS:

a. : Injection doesn't create any flowpath thru the core,

c. BOP busses 1A,1B,10 are deenergized,

d. Doesn't occur until substantial RCS inventory lost (not expected in this case).

1-l.

(-

1987 1988 Annual Requalification Examination Part B (Answer Key)

QUESTION: OR-0358 TIME: 1 MIN POINT VALUE:LO A reactor trip has occurred. The reactor trip and bypass breakers are open and neutron flux is decreasing. Rod bottom lights are on except for rods D4 and J2. They indicate full out. Select the appropriate operator action with respect to core reactivity, a.

Borate using normal boration for 17 minutes for each stuck rod.

b.

Borate using normal boration for 17 minutes for the second stuck rod.

c.

Borate using emergency boration for 17 minutes for each stuck rod.

I d.

Borate using emergency boration for 17 minutes for the second i

stuck rod.

ANSWER:

c.(1.0)

COMMENTS:

T.S. requires > 301 Opm of > 7000 ppm boron and EOP-1,1 requires borating for each stuck rod. This is an area not covered specifically in the AOP and therefore falls under objective 9. The student must go to EOP-1.1 not 1.0 to find actions for all rod bottom lights not lit, l.

L l

L 7

i' 1987 1988 Annual Requalification Examination Part B (Answer Key)

QUESTION: OR-0370 TIME: 1 MIN POINT VALUE-100 hours after the plant shutdown, the RCS is at 140'F when a loss of the 2 vesseW:p" abovehe centerline and the vessel head is installed. H 4

much time does the operating crew have to restore RHR before the core begins to uncover?

ANSWER:

75 i 5 minutes COMMENTS:

Requires the examinee to recognize he is in a loss of RHR (EOP 2.4) and application of Att. II. Tolerance is given to 1/2 the smallest division.

4 b

p-y l

I-l.

I I

l

w 1987 1988 Annual Requalification Examination Part B (Answer Key)

QUESTION: OR 0371 TIME: 5 MIN POINT VALUE:1.0 The control room operators are performing EOP-4.2, SGTR with Loss of Reactor Coolant - S'2bcooled Recovery Desired. Offsite power is not available. SI has been 1

terminated and normal charging established per Step 18 but subsequently, the operators have determined that limited SI!1ow must be re Initiated to maintain i

adequate RCS inventory.

What is the proper way to re initiate SI flow?

1 a.

Cycle the reactor trip breakers to break the seal in on the SI reset. This permits another auto SI if needed.

b.

Manuallyinitiate SI.

i Manually start the second charging pump, c.

d.

Reopen HI HEAD to COLD LEG INJ valves MVG 8801A & B.

ANSWER:

d.

(1.0)

COMMENTS:

a,b results in undesired CNTMT ISOL, which complicates plant control.

c. Minimal effect w/ normal charging

d. Per step 22 alternate action, requires interpretation that " inability to maintain PZR level"is related to RCS inventory maintenance.

)

l l

L i

t l

t 1987 1988 Annual Requalification Examination Part B (Answer Key)

QUESTION: OR 0375 TIME: 5 MIN POINT VALUE:1.0 Offsite power is lost without SI actuation. The control room operators verify a reactor trip and a turbine trip. They determine that the D/Gs have re energized IDA and IDB. All appropriate loads have sequenced on. While ensuring that the RCS stabilizes at no load conditions, an operator observes that PZR pressure is 1980 psig

)

and slowly decreasing. He checks the PZR PORVs and spray valves and notes that all are closed. PZR levelis stable at 25%.

What corrective action,1f any, should be taken?

a.

No operator action is necessary. Pressure will stabilize at approximately 1850 psig.

b.

Manually actuate SI and return to EOP 1.0, Reactor Trip or SafetyInjection Actuation, Initiate a rapid depressurization to minimize RCS inventory c.

loss through the RCP seals.

d.

Reset NON-ESF LCKOUTS, then energize PZR Backup Heaters as necessary.

ANSWER:

d.

(1.0) i COMMENTS:

a. No indication of this. Pressure is already below normal post trip level,

b. Not req'd. No indication ofloss of coolant

c. Only appropriate for loss of all AC in this case charging and/or CCW cools seals.

DO NOT USE WITH OR 0374

i i

1987 1988 Annual Requalification Examination Part B (Answer Key)

I QUESTION: OR 0386 TIME: EMIN POINT VALUE:14 A large break LOCA occurred with multiple ECCS malfunctions. The control room operators havejust completed responding to a Solid Path condition on the Core Cooling status tree when they identify a solid path on the Containment status tree. Containment pressure is 62 psig and increasing.

I The operating crew must prevent both further steam pressure increases and Hydrogen ignition, resulting in a pressure spike.

l a.

b.

A sodium hydroxide reaction with the containmentliner.

l Containment flooding and subsequent equipment damage, c.

i

't d.

Containment overheating and subsequent equipment damage.

ANSWER:

a.

(1.0) i COMMENTS:

b. Not expected. (NaOH reaction with aluminum is a hydrogen source).

c, d. Not indicated /less important than overpressure.

i 1

I i

i 1987 1988 Annual Requalification Examination Part B (Answer Key)

QUESTION: OR-0392 TIME: 6 MIN POINT VALUE:1.0 Component Cooling pump "A"is running, aligned to the active loop.

j Component Cooling pump "C"is aligned to the inactive loop for performance ofits quarterly pump STP (122.002). Prior to starting pump "C", a safety injection occurs. (Lineup prior to SI: "A" pump running,"B" pump off with breaker racked up,"C" pump to "B" train, off with breaker racked up). Which component cooling pumps will be running following the safety injection, assuming no operator i

action?

(1.0)

A.

"A" pump only B.

"B" pump only O.

"A" and "B" pumps only i

D.

"A", "B", and "C" pumps ANSWER:

A.

(1.0)

COMMENTS:

)

Interlock prevents starting either CCW pump on "B" Train therefore, only "A" pump 2

will start, i

l 4

j 1987 1988 Annual Requalification Examination Part B (Answer Key)

QUESTION: OR-0404 TIME: 4 MIN POINT VALUE:LO Recovery from a 500 gpm tube rupture is in progress per the EOPs. A rapid RCS cooldown hasjust been completed and the steam dump controller has been taken to manual-close. To what setpoint should the steam dump controller be adjusted if the following conditions exist:

(1.0) i e

Ruptured S/G pressure-1100 psig e

Core exit thermocouple temperature - 535'F e

Steam header pressure - 845 psig i

e RCS pressure 1645 psig A.

4.1 i

B.

6.5 i

r C.

8.4 D.

10.0 l

ANSWER:

i B.

(1,0)

COMMENTS:

Requires determination that steam dumps should be set to maintain current steam header pressure, and calculation of appropriate setpoint (845 + 1300 = 0.650) s

r l

1987 1988 Annual Requalification Examination i

Part B (Answer Key) i QUESTION: OR-0408 TIME: 4 MIN POINT VALUE:1.0 Recovery from a 500 gpm tube rupture is in progress per EOP-4.0. Just prior to the rapid RCS cooldown, note 17(3) directs the operator to take both STMLN SI switches to block when the P-12 status light becomes bright. Failure to perform this operation at this specific point in EOP-4.0 i

could result in which of the following during the subsequent cooldown?(1.0)

{

A.

Main SteamIsolation B.

Automatic reinitiation of SI C.

Turbine Driven EFW pump trip D.

Main Steam Isolation and Automatic Reinitiation of SI i

ANSWER:

A.

(1.0)

COMMENTS:

B, D SI reset has been locked in by P-4 C. No TDEFW Pp trip on low pressure P

,Q.

t vv-.-

w w

+-,w--

1 vw a

.w us v-v-

1987 1988 Annual Requalification Examination Part B (Answer Key) l QUESTION: OR-0416 TIME: 5 MIN POINT VALUE:1.0 During a normal surveillance, the following RWST parameters were recorded:

t e

Volume 455,000 gal.

o Temperature - 68'F t

e Boron Concentration - 1950 ppm Corrective action must be taken because (1.0)

A.

Boric acid may crystallize, blocking ECCS flow B.

The reactor may not remain suberitical following an analyzed accident.

}

C.

Not enough water is available for both injection and RB spray flows.

l t

L D.

Containment flooding could occur due to sump t

overfill during a LOCA.

t ANSWER:

B.

I COMMENTS:

A is incorrect since boron precipitation is not a concern at this

. tem perature and concentration. C is incorrect since this is a more than suff cient water volume for SI & spray flow. D is incorrect since

. flooding is not a concern during a loss of coolant accident.

DO NOT USE WITH OR-0379 1.

I 1

• l

)

1987 1988 Annual Requalification Examination Part B (Answer Key) j QUESTION: OR-0429 TIME: 1 MIN POINT VALUE:LQ The reactor is at power with all systems operating normally. Control bank D is at 230 steps. Control rod H6 drops; the reactor does not trip.

1

. How should the negative reactivity from the dropped rod be balanced out? (Assume a negative MTC).

I Permit the plant to cooldown, adding positive reactivity.

i a.

b.

Alternate dilute the boron concentration to add positive reactivity.

c.

Reduce power so that lowering the equilibrium Xenon concentration adds positive reactivity.

.i d.

Reduce power so that lowering the power defect adds positive reactivity.

. ANSWER:

I

'd.

COMMENTS:

a. Required cooldown would be excessive.

b. Too slow to prevent excessive C/D

c. Xenon too slow; also, Xenon will Initially increase on downpower.

d. Requires interpretation of basic of AOP-403.6 step.

+

t h

x- - - - - - -. _ - - - _ -. _ _ _ _ - __ _ _ _ _ ~ - -.. _.. - -, - - -,,,

o

]

1987 1988 Annual Requalification Examination Part B (Answer Key)

QUESTION: OR-0430 TIME: 1 MIN POINT VALUE:LQ A reactor trip occurs from full power. The control room operators verify i

that the reactor and turbine are both tripped and that the ac emergency j

buses are both energized. An operator checks the SI first out annunciator and the ESFIE status lights, none of which is illuminated. He checks primary pressure. It is 1845 psig and steadily decreasing.

For this reactor trip, SI has a.

Not occurred and is not required.

b.

Notoccurred butis required.

c.

Occurred andis required.

i

.i d.

Occurred butis not required.

ANSWER:

b.

COMMENTS:

. Requires application of EOP-1.0 symptoms 11.3 to a specific case.

t 1

F 4

1987 1988 Annual Requalification Examination Part B (Answer Key)

{

QUESTION: OR-0432 TIME: 3 MIN POINT VALUE:LQ The reactor trips from 100% power with SI actuation. The cause is a steamline break downstream of the MSIVs. The control room operators L

take actions as per EOP 3.0 and obtain the required plant / system / component responses, except for the following: the MSIV associated with S/G A

)

will not shut, and the maximum total EFW flow rate that can be achieved is

)

300 rpm.

i S/G levels are as follows:

i S/G A:

offscale low, narrow range S/G C:

45% narrow range S/G B:

ofTscale low, narrow range Which of the following correctly describes the impact of this configuration on j

secondary heat removal capability?

1 a.

Secondac heat removal capability has been lost because EFW flow is inadequate.

b.

Secondary heat removal capability has been lost because S/G levels are inadequate.

{

c.

The loss of secondary heat removal capability is imminent, unless S/G Aisisolated.

?

d.

Secondary heat removal capability is adequate at current S/0 C levee

.y ANSWER:

d.

COMMENTS:

Requires interpretation of basis of step to decide that S/G C level is adequate for heat removal.

l

"i o

1987 1988 Annual Requalification Examination Part B (Answer Key)

QUESTION: OR.0434 TIME: 1 MIN POINT VALUE:LO j

The reactor trips from 100% power due to a spurious signal from the reactor protection system. The control room operators verify that the reactor is tripped and immediately note that the turbine is not tripped.

Because of a malfunction in the turbine trip system, none of the turbine stop, control, or combined intercept valves have closed.

Ifleft uncorrected, this malfunction is likely to result in a.

The main turbine overspeeding, possible resulting in damage to the main turbine rotor and shaft, b.

A loss of condenser vacuum, resulting in the loss of condenser i

steam dumps.

.i

.i l

t c.

An increase in RCS pressure, possibly resulting in the PZR 1

PORVs lifting, d.

An uncontrolled cooldown of the RCS, resulting in reduced shutdown margin.

ANSWER:

d.

COMMENTS:

a. Should be stopped by speed error on control & intercept vlvs: also, generator has not tnpped,

b. Not expected w/ cire, water pumps running.

c. Overcooling will decrease RCS pressure, j

t