IR 05000293/1987056
| ML20148B977 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 03/08/1988 |
| From: | Howe A, Lange D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20148B948 | List: |
| References | |
| 50-293-87-56OL, NUDOCS 8803220221 | |
| Download: ML20148B977 (142) | |
Text
{{#Wiki_filter:U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO.
87-56 (OL) FACILITY DOCKET NO.
50-293 FACILITY LICENSE N0.
DPR-35 LICENSEE: Boston Edison Company M/C Nuclear 800 Boyleston Street Boston, Massschusetts FACILITY: Pilgrim Nuclear Power Station EXAMINATION DATES: Decembe-7 - 11, 1987 CHIEF EXAMINER: M d.
- ,2.,3.5 -88 A. Howe, Senior Operations Engineer Date APPROVED BY: _ _ W 3-P-PP 0. Lange, Chie, BWR Sedtion, Date Operations Branch, DRS SUMMARY: Written examinations and operating tests were administered to two senior reactor oporator (SRO) and six reactor operator (RO) candidates.
All candidates passed these examinations.
8803220221 080315 PDR ADOCK 05000293 V PDR f
DETAILS TYPE OF EXAMINATIONS: Replacement EXAMINATION RESULTS: R0 SRO - Pass / Fail [ Pass / Fail [ ~ Written _ 6/0 [ 2/0 [ - [0perating [ 6/0 [ 2/0 [ ~ - 6/0 [ 2/0 _ - _0verall _ _ _ _ _ 1.
CHIEF EXAMINER AT SITE: A. Howe, Senior Operations Engineer 2.
OTHER EXAMINERS: R. Turner, Operations Engineer C. Gratton, Reactor Engineer, OLB D. Moon, Battelle, PNL 3.
The following is a summary of generic strengths or deficiencies noted on operating tests.
This information is being provided to aid the licensee in upgrading license and requalification training programs.
No licensee response is required.
_ _ - - - _ __.
) .
STRENGTHS a.
Initial' event diagnosis.
b.
General knowledge of procedures, i.e. surveillance tests, plant startup procedure, routine operations, c.
Use of alarm response procedures, d.
General knowledge of administrative procedures.
DEFICIENCIES a.
Ability to quickly locate abnormal procedures.
4.
The following is a summary of generic strengths or deficiencies noted from the grading of written examinations.
This information is being provided to aid the licensee in upgrading license and requalification training programs.
No licensee response is required.
REACTOR OPERATOR STRENGTHS a.
Understanding of delayed neutron effects on reactor operations.
b.
Understanding of reactivity coefficients, c.
- Knowledge of the Full Core Display.
d.
Understandingsof the effects of the installation of the shorting links on the reactor protection system.
e.
Knowledge of the operation of the Rod Block Monitor system.
f.
Knowledge of whole body exposure limits.
DEFICIENCIES a.
Knowledge of refueling bridge interlocks.
b.
Knowledge of automatic and manual scram bypasses, c.
Understanding of reactor level instrumentation systems.
- _ . -. ~. .-
, SENIOR REACTOR OPERATOR STRENGTHS < a.
Knowledge of thermodynamics and fluid flow, b.
Understanding of Keff and period.
Knowledge of Low Pressure Coolant Injection loop select logic.
c.
d.
Knowledge of the Standby Liquid Control (SLC) sys+.em.
Knowledge of symptoms for small leaks in the drywell and symptoms e.
for a failed jet pump.
WEAKNESSES Understanding of power shaping control rods.
a.
b.
Erroneous knowledge that securing all low pressure ECCS pumps will cause the Automatic Depressurization System timer to reset, Knowledge that a heater failure will cause the Standby Gas c.
Treatment fan to trip.
.d.
Knowledge that flow biased rod blocks will occur on a failure of a flow converter, Knowledge of E0P-2 bases for inhibiting ADS when injecting SLC.
e.
5.
Personnel Present at Exit Interview: NRC Personnel A. Howe, Senior Operations Engineer, USNRC C. Gratton, Reactor Engineer, USNRC Facility Personnel P. Mastrangelo, Chief Operating Engineer, BECO J.' Alexander, Operations Manager, BEC0 H. Balfour, Operations Training Group Leader, BECO E. Ziemanski, Nuclear Training Manager, BECO L. Beckwith, Compliance, BECO T. Sullivan, Watch Engineer, BECO K. Highfill, Station Director, BECO P. Freehill, Gulf States Utilities
, _ _ - _ _ _. _ _ _._. _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _
6.
Summary of NRC comments made at exit interview: The chief examiner thanked the training and operations staffs for < their cooperation. The written exam review went well with a few initial comments identified.
- . , The operating test generic strengths and weaknesses were discussed. The chief examiner also noted that this group of candidates demonstrated an overall better concept of plant operation and integrated plant response when compared to the previous license class. The apparent reason for this - improvement was the use of the Pilgrim plant specific simulator as a part
of their training.
, The chief examiner requested that the simulator be used extensively to prepare the licensed operators for transients prior to the startup and startup test program. This use of the simulator should also include the operators with limited licenses as active participants. Also these operators with limited licenses should observe the performance of testing on the plant, , 7.
Summary of facility comments and commitments made at exit interview:
The facility staff committed to use the simulator to train operators for startup testing, s Shortly after the exit meeting comments were completed and in response to the weakness identified on the operating tests, the Training Department committed to revise the procedure index to make it easier to use.
I Attachments: 1.
Written Examination and Answer Key (RO) 2.
Written Examination and Answer Key (SRO) "
3.
Facility Comments on Written Examinations after Facility Review ' 4.
NRC Response to Facility Comments ! , & P r - r I .-
. A Haa-c e. I a wy "= g , _ , .. U.
S.. NUCLEAR REGULATORY COMMISSION-REACTOR OPERATOR LICENSE EXAMINATION FACILIT": _PII; GRIM _________,_ ,,__ REACTOR TYPE: _HWR;@g3___________,_____ DATE ADMINISTERED: _@7flgfg@_'______________ , . EXAMINER: _MggN3,93________________ CANDIDATE: ____ l _____________ INSISUCI]QNg_IQ_C9Ngip91El Une separate paper for the answers.
Write answers on one side oniv.
Staole cuestion sheet on too of the answer sheets.
Points'for each' auestten are indicated in parentheses atter the cuestion.
The passing grade reautres at least 70% in each category and a final grade of at least 50%. Examination papers wt il be'otcked uo six (6) hours after the examination starts.
% OF RTEGORY % OF CANDIDATE'S CATEGORY mye(UE, _1g196 ___SCgSE___ _y@6UE__ ______________C@lEGQS!_____________ 3@199-- 2Et99 l.
PRINCIPLES OF NUCLEAR POWER
. ______ PLANT OPERATION. THERMODYNAMICS.
HEAT TRANSFER AND FLUID FLOW 3Ez99-- 2Ez99 2.
PLANT DESIGN INCLUDING SAFETY ___________ ________ AND EMERGENCY SYSTEMS -_-----____ ________ 3.
INSTRUMENTS AND CONTROLS _ 3Ez99-- 2E199 2EA99-- 2Ergg
PROCEDURES . NORMAL. ABNORMAL, __,________ ________ EMERGENCY AND RADIOLOGICAL CONTROL 99299-. ________% Totals ___________ Final Grade All work done on thi s ex aminati on is my own.
I have neither given nor received aid.
Candidate's Signature . .
. -. _ _ _ _ _ - _ - _ _ _ - _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _,. _ _ _. _ ' s .. . NRC RULES AND GUIDELINES FOR LICENSE EX4MINATIONS i 6 ring the administration of this examination the followirg rules apolvt F Cheating on the examination means an automatic denial of your application
> and could result in more severe penalties.
, Restroom trios are to be limited and only one candidate,at a time may > leave.
You must avoid all contacts with anvene outside the examination room to avoid even the appvarance or possibility of cheating.
i Use olack ink or dark cencil only to facilitate legible reproductions.
, Print vour name in the blank crovided on the cover sheet of the
! examination.
i Fill in the date on the cover sheet of the examination (if necessary).
> i f { Use only the paper orovided for answers.
. Print your name in the upper right-hand corner of the first page of each . . section of the answer sheet.
Consecutive 1v number each answer sheet. write "End of Categorv __" as
appropriate, start each category on a new page, write gniv on gne side i ' of the paper, and write "Last Page" on the last answer sheet.
L Number each answer as to category and number, for example, 1.4, 6.3.
j , ' 0.
Skip at least thtee lines between each answer.
f l.
Separate answer sheets from pad and place finisned answer sheets face down on vour desk or table.
, l 2. Use abbreviations only i f thev are commonly used in facility litetatute.
3.
The coint value for each cuestion is indicated in carentheses after the question and can be used as a guide for the death of answer required.
,
Show all calculations, methods. or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
r i 5.
Partial credit may be given.
Therefore. ANSWER ALL PARTS OF THE t QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
[ 6.
If parts of the examination are not clear as to intent, ask questions of
the ex ami net 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 examinaticn has been completed.
. e
_ __ . . . . D. When you complete your examination, you shall: a.
Assemble your examination as follows:
(1) Exam cuestions on top.
(2) Exam aids - figures. tables, etc.
, (3) Answer pages i n c l'ud i n g figures which are part of the answ . b.
Turn in your copy of the afxamination and all pages used to answer the ex ami n at i on cuestions, c.
Turn in all scrao paper and the balance of the paper that you did not use for answering the cuestions.
d.
Leave the examination area. as defined by the examiner.
If after leaving. vou are found in this area while the examination is still in progress. vour license may be denied or revoked.
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. . 11._ESIUC1ELES_gE_ NUCLE 98_EgWEB_E(9NI_QEE8911gN.
PAGE
THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW . QUESTION 1.01 (2.50) a.
DOES the delayed neutron fraction INCREASE or DECRE,ASE from the beginning of cycle (BOC) to the end of cvCle (EOC)' (0.5) b.
WHAT is the major cause for the change in delaved neutron f rac ti on from BOC to EOC? (1.0) c.
HOW do delaved neutrons affect reactor coerations such that control of reactor power is possi bl e? (1.0) OUESTION 1.02 (2.00) For each of the following events. STATE which COEFFICIENT of rocctivity (MODERATOR. VOID. DOPPLER) would act FIRST to chance FOCCtivity.
c.
control rod drop at 25 percent power (0.5) b.
SRV opening at 50 percent power (0.5) c.
loss of shutdown cooling when removing decay heat (0.5) d.
one recirc pumo trips while at 50 percent power (0.5) QUESTION 1.03 (3.00) For the f ollowing changes in plant parameters WILL control rod worth INCREASE. DECREASE or NOT BE AFFECTED? Briefly EXPLAIN WHY? e.
An increase in moderator temperature.
(1.0) b.
An increase in void content.
(1.0) c.
An increase in fuel temperature.
(1.0) . ' (***** CATEGORY 01 CONTINUED ON NEXT PAGE
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t l THERMODYK4MICS. HEAT TRANSEER,@ND_ELU1D_ELQW , t
J OUESTION .l.04 (2.00) , The reactrar'has Deen coerating at 95 percent power.for s,ev er a l
,
devs.
An coerator RAPIDLY reduces reactor oower to 60 percent by reducing the Opeed of the recirculation pumps.
During the
, ' next'2-3 MINUTES the coerator notices that the reactor oower-slowly increases ',*.o 63 percent (with no coerator action).
- EXPLAIN the cause of the power increase.
(2.0) QUESTION 1.05 (3.00) i i a.
Reactor power has increased from 15 on IRM range 1 to 15 on IRM range 3 in 180 seconds.
The point of s adding heat.is determined to be 25 on IRM range 8.
HOW much longer will it take f or reactor power to
reach the coint of. adding heat if reactor period , remains constant? SHOW all work.
(2.0) ] , b.- During a rapid power increase, very short reactor periods can be maintained. yet for rapid power decreases, the - stabl e reactor period is limited to -80 seconds.
EXPLAIN ' -the reason for this difference.
(1.0) , QUESTION 1.06 (1.50) , Assume the reactor is at 100 percent power and the xenon has built up to 3 percent dk/k - a.
If the reactor scrammed. HOW long would it take for the - l xenon concentration to peak? (C.75)
I b.
HOW many hours would it take for the xenon to decay to approximately zero percent dk/k assuming the reactor . remains shutdown? (0.75) f
> I ! I E ,
b e '
,
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t THERMODYNAMICS. HEAL TRANSFER AND FLUID FLOW .. QUESTION 1.07 (3.00) Concernino core thermal l i mi ts: , c.
Specifically WHAT adve'rse physical consecuence could occur in the reactor vessel if the MAPRAT limit is exceeded? (1.0) b.
For each of the following conditions. STATE whether the critical oower level would INCREASE. DECREASE. or REMAIN UNCHANGGD.
Brief1v EXPLAIN WHY, 1.
Extraction steam to a high pressure feedwater heater isolates.
(1.0) , 2.
a recirculation pumo speed reduction using '.he M/A transfer station.
(1.0) Ol'TS T I ON 1.08 (1.00) Tho reactor is operating at a steady state oower tevel of 75 pcreent.
A central control rod is withdrawn from notch position 44 to position 48 and reactor power is observed to d: crease slightly.
A senior reactor operator informs you tho power decrease was due to the reverse power effect.
DESCRIBE WHAT is happening in the core to cause this phgnomenon.
(1.0) . QUESTION 1.09 (2.00) WHAT are four (4) plant or component design features or opcrational limitations which ensure adequate net positive cuction head for the reactor recirculation pumos? (2.0) . ' (***** CATEGORY 01 CONTINUED ON NEXT PAGE
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y THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW .. QUESTION 1.10 (3.00) DS:cribe HOW (INCREASE or DECREASE) and WHY the discharg,e head of cn RHR pump is affected for each of the following.
(Consider each condition separately and as'sume NPSH is maintained in all cases.)
c.
Suction oressure increases.
(1.0) b.
The discharge valve is throttled closed.
(1.0) c.
The temoerature of the fluid being pumoed increases.
(1.0) OUESTION 1.11 (2.00) e.
Most condensers are designed with excess condensino capabilityt that 15.
the condensed l i Qui d leaves the condenser hotwell several degrees below the saturation temperature.
HOW would PLANT EFFICIENCY be affected (INCREASE. DECREASE.
or NOT AFFECTED) if the temperature of the circulating water was greativ DECREASED? EXPLAIN your answer.
(1.0) b.
If the main condenser was absolutely air tight. WOULD there be any need for the air ejectors? Explain WHY, (1.0) . ' (***** END OF CATEGORY 01
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. . EL__C6@UI_QESl@U_iGq(UglGQ_@@EEI!_@yg_E[gSQENgy_@y@IENS PAGE o OUESTION 2.01 (2.00) For the Rod Block Monitor (RBM). PROVIDE answers to the following aucetions: , . 3.
WHAT adverse condition is the system det,1gned to prevent? (1.0) b.
When the Meter Function Switch on the Back Panel 937 Meter Section is in the "Count" cosition. WHAT are the "units" of the indication on f' cter and WHAT can be calculated by utilizing the indit value? (1.0) QUESTION 2.02 (3.00) For eacn of the following situations. DETERMI' "hether or not the activity can occur.
If the activity can NOT x ur. WHAT must change to allow it to occur.
c.
Refuel bridge i s over the reactor vessel and in motion toward the fuel pool with the fuel grapple loaded.
All rods are inserted.
The reactor mode switch position is changed from REFUEL to STARTUP.
WILL the bridge continue to moce? (1.0) b.
Refuel platform is over the vessel.
The frame mounted hoist is loaded.
One rod is at position 30.
CAN the load on the hoist be lowered into the vessel? (1.0) c.
Refuel platform is over the vessel with the mode switch in REFUEL.
The grapple is fully lowered and unloaded.
CAN a control rod be withdrawn? (1.0)
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.. .OUESTION 2.03 (3.00) MATCH the appropri ate source ot cooling water to each of the 'following loads: (3.0) , Loads - Sources .-----------------------------------------
a.
high pressure service air compressor 1.
RBCCW b.
'HPCI room cooler 2.
TBCCW c.
drvwell air cooling coils 3.
seawater system d.
fuel pool cooling heat exchangers
salt service water system e.
RBCCW heat exchangers f.
control room HVAC refrigeration units g.
turbine lube oil coolers h.
condensate Iump motors i.
residual heat removal pump seal c ool ers j.
reactor feed pump seal water coolers , QUESTION 2.04 (2.50) Concerning the drywell leak detection system
. a.
A drywell equipment drain sump high level is annunciated in the control room.
If level continued to increase. WHAT three (3) other signals or acticns occur as a direct result L of a HIGH LEVEL and HIGH-HIGH LEVEL? (1.5) b.
The drywell equipment drain sumos and the drywell H2/02 , Monitoring system isolate on a Group II PCIS signal.
' DESCRIBE WHAT operator action can be taken, if any, to place the H2/02 monitor back in service before the Group II PCIS signal is cleared.
(1.0) ! { ' l . (***** CATEGORY O2 CONTINUED ON NEXT PAGE
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c . 2 __E69NI_95 SIGN _1NCLUQ1NG_S9EgIy_9NQ_gDg69ENCy_gygIgDS PAGE
OUESTION 2.05 (2.00) With regard to the Full Core Displav: a.
WHAT two (2) signals will cause an AMBER light tc ' illuminate? - (1.0) b.
Each control rod has a BLUE light on the canel.
This light will illuminate when the ___,,______ valve and the __________ valve are both ___________ (valve position).
(1.0) QUESTION 2,06 (3.00) While the reactor is operating at 100 percent power. a comolete loss of essential instrument air e,c c u r s. Assuming no coerator action. DEECRIBE HOW and WHY the following parameters will chance prior to a reactor scram.
(3.0) a.
main condenser vacuum b.
RBCCW temperature c.
CRD cooling water flow d.
FW flow rate to the reactor vessel o.
FW temperature f.
indicated SLC tank level l i ,
l l l
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f?A__EL@NI_QgS]QN_JNCLUp]NG_@@ggly_9NQ_gMggggNCy_gy@IEMS .. QUESTION 2,07 (2.50) a., STATE the normaliand alternate power supplies to the Reactor . Protection System (RPS) bus.
(1.0) , b.
WHICH of the followind components would be directiv affected by a manual transfer of RPS bus A from its normal to alternate oower supply? (1.5)
IRM A .. 2.
APRM B 3.
reactor building vent i l at i on radiation monitors ~4.
off gas system radiation monitors 5.
.MSIV 6.
main steam line radiation monitors QUESTION 2.08 (3.00) The RCIC System started on an automatic initiation signal and has ~bsen operating for 5 minutes.
For each of the following conditions, STATE whether the RCIC system WILL or WILL NOT continue to operate.
If it will continue to operate. WILL there be any adverse effects from RCIC operation under the conditions? If it WILL NOT continue to operate. WHY NOT? Consider each condition separ at el y.
Assume r.o operator action, a.
The condensate pump for the barometric condenser fails cauri ng a high level in tne barometric condenser.
(1.0) b.
Reactor pressure decreases to 100 psig.
(1.0) c.
The RCIC lube oil pump f ails causing oil pressure to drop to 1-psig.
(1.0) . ' (***** CATEGORY O2 CONTINUED ON NEXT PAGE *****) - .. . _. - - -, - _.
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.. QUESTION 2.09 (1.50) cA core spray line breaks inside the shroud.
~ a.
'WILL the break cause an alarm in the control room? (O. 5). . . core sprav performance for b.
HOW will the break affect J ~ ' that loop? (1.0) QUESTION 2.10 (2.50) Concerning the Standby' Gas Treatment System (SBGTS): a.
~When an initiation signal occuts. the damoers to the exhaust plenum open.
Air is drawn from several locations.
Refer to the attached SBGTS Figure 1 and' LIST the three (3) locations-from which. air can be drawn which corresponds to the blanks on the Figure l abel l ed I, 2.
and 3.
(1.5) b.
.If'the SBGTS was automatically initiated by a reactor low , water level signal. WHAT must be done to shut the system down once the l ow-l evel signal clears? (1.0) . i l i.
l l l
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.. QUESTION 3.01-( 3. 00 ) Following_a. reactor SCRAM. some scram signals are bypassed by operator or automatic actions..For each of the f ol l owi n,g scram signals. STATE all'the condition (s) that must be in effect for a bypass to occur: ' (3.0) a.
main steam line isolation scram b.- reactor mode switch in SHUTDOWN scram c.
turbine control valve fast-clesure scram d.
scram discharge volume high l evel scram . QUESTION 3.02 (1.50) During shift turnover you are informed that the RPS "shorting links"'are scheduled to be installed during your shift.
WHAT two '(2) effects will i nstal l ati on of the "shorting links" have on the nautron monitoring trips and the reactor protection system? (1.5) i i
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.. .OUESTION 3.03 (2.50) For the following situations. STATE'whether the Automatic Depressurization System (ADS) relief valves will;OPEN.
C, LOSE or REMAIN AS IS.
Consider each set of conditions separately.
a.
-ADS initiating signal (low ' level) occurs, the ADS timer times out, and ADS valves open.
Then reactor water level rises to'lO~ inches.
(0.5) b.
ADS initiating parameters are present and ADS valves open.
Then the ADS timer reset buttons are'deoressed.
(0.5) c.
ADS initiating parameters are present and ADS valves open.
Then a DC cower failure occurs that affects'all busses suoplying ADS valves.
(0,5) d.
ADS initiating parameters are present, a loss of instrument air supply to the drvwell has occurred. and the 120-second timer i s timing out.
Then the 120-second timer times out.
(0.5) e.
ADS initiating parameters are present, all ECCS pumps are secured except for CS pump B which is running with a discharge pressure of 165 psig, and the 120-second timer is s ti mi ng ' out.
Then the 120-second timer times out.
(0.5) . . ' (***** CATEGORY 03 CONTINUED ON NEXT PAGE
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'13 - k .. i QUESTION 3.04 (2.00) Consider the reactor. water level instrumentation: a.
For each of the:following parameter' changes and coerational conditions. STATE whet'her the INDICATED vessel level will INCREASE, DECREASE. or REMAIN THE SAME for theisoecified level instrument.
ACTUAL vessel level REMAINS THE SAME.
- . 1.
The reactor vessel temoerature increas7s from 130 degrees F to 200 degrees F during a reactor startup.
HOW Will the SHUTDOWN RANGE level instrumentation respond? (0.5) 2.
The reactor is in cold shutdown.
The reactor recirculation pumps trip.
HOW will the FUEL ZONE l evel instrumentation respond? (0.5) .b.
WHICH water level instrument RANGE provides the low water level trip inputs to the RPS logic? (0.5) c.
WHICH water l evel instrument RANGE provides the water l evel inputs to RHR Centainment Spray logic? (0.5) QUESTION 3.05 (2.00) The reactor is operating at 70 percent power.
Flow convertor A ' fails such that its output is downscale.
STATE ALL the CAUSE(s) t A ( g yy. red * 5 M /b Nrx8.
(2.O) for each trip which wi11 occur.
okhL& W* $t(. N @, , QUESTION 3.06 (3.00) Consider the Rod Block Monitor system (RBM): a.
WHAT i s the purpose of the null sequence control circuit? (1.5) b.
WHAT are three (3) ways the RBM trips can be BYPASSED? (1.5) l l . ~ (***** CATEGORY 03 CONTINUED ON NEXT PAGE
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i .. ' QUESTION 3.07 ( 3. 00 ) ' Consider the Process Radiation Monitoring (PRM) System: a.
Specifically. WHERE. in relation to the MSIVs. are "the ' main steam line (MSL) ' radiation monitoring detectors located? (1.0) 6.
On a trio of the MSL radiation monitors. WHAT three (3) automatic acticos, besides the closure of the MSIV's, could occur as a direct result of the MSL Hi Hi Rad si gnal ? (1.5) c.
WHAT. if any, automatic action (s) occurs if one channel of off-gas PRM fails high while the other channel of post-treatment offgas PRM is downscale? (0,5) NHUtnU$ ps QUESTION 3.08 (2.00) The reactor is operating at 100 percent power with the following Machanical Hydraulic Control (MHC) System setpoints: EPR setpoint: 930 psig MPR setpoint: 933 psig Bypass Opening Jack setpoint: O percent (BPV's are closed) Speed / Load Changer demand: 100 percent (rated load position) Briefly DESCRIBE the MHC system response to the following two ssparate situations.
INCLUDE in your discussion any changes in reactor pressure, control valve position and bypass valve position and WHY these changes occur.
Assume no operator ac t i r,n. a.
The EPR fails so that reactor pressure decreases.
(1.0) l b.
Power is lost to the EPR.
(1.0) l l l l , l . ' (***** CATEGORY 03 CONTINUED ON NEXT PAGE
- )
i --- , .- - .,,,m ,, _..,.. -, ,,-m,---. - - - - . - - - -,,. ~,,. _ _ _ _.., - - - - ~.... _.. _. _ -, _,. -.. _., - _,,,. - -
. l 3.
INSTRUMENTS AND CONTROLS PAGE
QUESTION 3.09 (3.00) The reactor is operating at 100 percent power with the encirculation M/A transfer stations in MANUAL.
Explain HOW and WHY EACH recirculation cumo responds to the f ol l'owi no ^ conditions.
Where applicable. PROVIDE soecific values.
a.
Reactor water level decreases to 17 inches following a feedwater ouma trio.
(1.0) b.
Full open signal from recirculation pumo A discharge va4.e is lost.
(1.0) c.
The feedwater-f l ow signal to the recirculation system f ails to zero.
(1.0) OL'iST I ON 3.10 (3.00) The reactor is operating at 100 percent power under steady state :ondi tir l s.
For the following loss of signal failures to the 9eedwater control system, INDICATE in which direction r@ccter water level will respond (INCREASE or DECREASE). the REASON for the change, and any probable AUTOMATIC ACTION that will occur directly as a result of the change in water level.
Assume no operator action.
(3.0) c.
Loss of one feedwater flow input b.
Complete loss of steam flow input c.
Loss of reactor water level input d.
Loss of density compensation to the GEMAC detectors i
- I
~ (***** END OF CATEGORY 03
- )
. $t__ESggEQU65@_;_UQBd@bt_@@Ug6096t_Ed@BQEdgy_@NQ PAGE
86919LgGigg(_ggN169L . QUESTION 4.01 (2.00) The reactor is operating at full power and the +cllowing, alarms occur: . PANEL C-7 TROUBLE ALARM and DRYWELL AIR COOLER HIGH ORAIN FLOW ALARM In accordance with Procedure 2.4.14 Leaks Insi de Pri mary Containment. LIST four (4) possible symotoms (other than alarms) that you would observe pri or to a reactor scram.
(2.0) QUESTION 4.02 (3.00) a.
According to Procedure 2.4.23.
"Jet Pump Flow Failure."
unexplained changes in jet Dumo flow could indicate jet pump failure.
WHAT are two (2) Other indications or symptoms of jet pump failure? (1.5) b.
WHAT two (2) actions would vou take upon observing the above symptoms? (1.5) DUESTION 4.03 (3.00) In accordance with Procedure 2.4.143.
"Shutdown from Outside the Control Room:" m.
WHAT are all the immediate actions that should be performed prior to evacuating the centrol room? (2.4) b.
WHAT three (3) systems could you use to control reactor pressure from outside the control room? (0.6) OUESTION 4.04 (1.50) While the reactor is at 50 percent power. an automatic initiation of the core spray system occurs during otherwise normal plant operation.
WHAT administrative or procedural conditions must be satisfied before you secure the core spray pumo? (1.5) , ' (***** CATEGORY 04 CONTINUED ON NEXT PAGE
- )
. - - .
-l .di__E69ggguggg_;_yg60963_gggggdg63_gDg5@gNgy_gyg PAGE
669196991996_G9NI696-l ,, QUESTION 4.05 (3.00) During a maintenance outace, a 23 vear old male worker wJth a lifetime exoosure of 23 REM (NRC Form 4 on file) is assigned to work in a 200 mrem /hr r a'd i a t i on area.
-The worker has received 250 mrem so far this calendar cuarter, a.
WHAT is-the maximum whole body exposure the worker could receive this quarter without exceeding the IOCFR2O allowable whole body exposure limits? (1.0) b.
HOW long canfthe worker be in the radiation area without exceeding PNPS administrative whole body occupational exoosure limits? SHOW vour work.
(1.5) c.
WHAT would the PNPS administrative wnole body occupa t i ona l exposure limit be for the worker's 18 year old brother who has no lifetime exposure (NRC Form 4 on file)? (0. 5 )- QUESTION 4.06 (2.00) While the reactor is operating at 50 percent power, a RCD DRIFT alarm occurs and you observe two Full Core Display drift lights ! are on.
.WHAT three (3) immediate operator actions are required according to Procedure 2.4.11, "Control Rod Drive Malfunctions?" (2.0) i OUESTION 4.07 (2.00) For each of the functions described below. MATCH it with the l appropriate tag that could be used.
(2.0) FUNCTION TAGS
..-------
a.
indicate that a piece of equipment is 1.
white tag with being tested green border b.
hung by the Nuclear Operating Supervisor 2.
vellow tag to prevent operation of a piece of equipment 3.
blue tag c.
provide visual indication that coeration 4.
orange tag l of a device is prohibited 5.
red tag d.
alert an operator that a piece of equip-ment is temporarily, operating not in accordance with design (***** CATEGORY 04 CONTINUED ON NEXT PAGE
- )
L --.- , ..., . -. - . -.. _ _. . - - . -- _
_ - .. _ - - .. .. . . - .. . .-.
. a fi__EBgCEgUBES_;_Ng80@L _9BNg8M8L _gDEB9ENCy_9Ng PAGE
1
.69919LggIC96_CgNI69L . - OUESTION 4.08 (3.00) a '. LIST the five (5) entrv conditions for EOP-03. "Primary ~ Containment Control.", INCLUDE setooints.
(2.5) b.
LIST the entry condition for EOP-05. "Radioactive-Release Control."
(C.5) QUESTION 4.09 (1.50) For each of the following situations STATE v&tich. if any. Emergency Operating Procedure (s) should be entered.
If none, state NONE.
.a.
Water l evel in HPCI pumo room floor is 2 inches deeo.
(0.5) b.. Reactor oressure reaches 1090 psig following a scram.
(0.5) c., Reactor building exhaust radiation level is 800 cos.
(0,5) , 0UESTION 4.10 (1.50) PROVIDE the f ollowing action levels or l i mi ts for plant operation: a.
Maximum allowable temperature difference between the vessel dome and bottom head drain just prior to starting a recirculation pump.
(0.25) b.
Maximum allowable temperature difference between recirculation loops just prior to starting a (0.25) recirculation pump.
c.
Maximum permitted cooldown rate during a normal shutdown.
(0.25) d.
Maximum pressure differential across an MSIV prior to opening the MSIV.
(0.25)
- e.
Maximum permitted RPV shell flange to shell temperature ' differential.
(0.25) f.
Maximum power level for mechanical vacuum pump operation.
(0.25) , . ' (****h CATEGORY 04 CONTINUED ON NEXT PAGE
- )
i ,,... ~..._.-__ _.._,-- -_ _ _,..___.,.__., _, _.,,._.. - _._ _ _ _.,_. ___ _..__,.,,_ . - _ _., _..... -_ .
. 9t__E8ggEDySES_ _UQ60@6t_@@UO60@6t_gdES@gdgy_9ND PAGE
869196901G66_GQN16g6 . QUESTION 4.11 (2.50) a.
Proceoure 2.1.3.
"Startup with MSIV's closeo 7x Pr e,s s u r e less than 600 psig" states that if it has been determined the reactor will remal'n in a hot standby condition for a period of greater than 8 hours, reactor pressure shall be decreased to 120 psia.
WHY is this action repuired? (1.0) b.
Procedure 2.1.5.
Section C.
"Operation After Reactor Scram with MSIV's Closed." states that reactor water level should , be maintained above 46 inches (above instrument zero) if both recirculation pumps have tripped.
WHY is the action reautred? (1.5) . ' (***** END OF CATEGORY 04
- )
(************* END OF EXAMINATION
- )
. MkSTF6 Ccty - TL__ PRINCIPLES OF NUCLEAR POWER PLANT OPERATION PAGE
t THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW .. ANSWERS -- PILGRIM-87/12/08-MOON, D.
' ANSWER 1.01-(2.50) , a.
decreases C+0.53
b.
The contribution to the delayed neutron oopulation by U-235 decreases as the U-235 is burned out and the contribution from plutonium increases, decreasing the delaved neutron fraction C+1.03.
c.
Delayed neutrons (make control of reactor power possible since the delayed neutrons make up part of the succeeding generation neutron population and) are.added to the population at a rate such that the reactor period can be controlled by the relative slow rod movement C+1.03.
REFERENCE 1.
Pilgrim Reactor Theory, pp. 3-29 through 3-33.
KA 292OO3K104 (2.4/2.4) ANSWER 1.02 (2.00) a.
Doppler b.
void c.
moderator d.
void C+O.53 each . REFERENCE 1.
Pilgrim Reactor Theory, pp. 7-2 through 7-23.
KA 292OO4K114 (3. d3.3) .. $ .
- . . . - . - . Li__EBINCIE(ES_QE, NUCLE @E_EgWEE_E(@NI_gEEE@I[QN PAGE
3 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW .. ANSWERS -- PILGRIM- -87/12/08-MOON.
D.
ANSWER 1.03 (3.00) , a.
As moderator'temperatu're increases the leakage of thermal neutrons from the fuel bundles into.the control rod regions increases C+0.53.
Thus rod worth i ncreases C+0.53.
b.
Moderator density decreases resulting i n more fast _ neutrons and fewer thermal neutrons leaving the bundl e C+0.53.
Since control rods are thermal absorbers. overall control rod worth decreases.C+0.53.
C s ab.- dixurwv. h a.M c% is,,:ta,4.J M i m,u m c htLK or k.4,a.Jf.a L anuale c6,c alss m4abe p.- L'l cad.
. i c.
Not affected C+0.53.. Since fuel temoerature affects crimarily AcN*' ,, Y /hkf 7 fast neutrons, which bre resonantly caotured, and control rods are thermal neutron absorcers, fuel temperature and rod worth ' are essentially independent of each other C+0.53.
REFERENCE 1.
Pilgrim Reactor Theorv. pp. 5-12 through 5-14 KA 292OOSK109 (2.5/2.6) ANSWER 1.04 (2.00) The reactor is now producing less steam to go to the turbine.
Thsr e will be less extraction steam (and reheater drain steam) going jf/p_, 2/ff87 to the feedwater heater C+1.03.
Therefore, less feedwater heating / j will occur resulting in colder feedwater entering the vessel C+0.53 which will cause reactor power to increase (about 3 percent) from the positive reactivity addition (alpha m) C+0.53.
REFERENCE 1.
Pilgrim: HT&FF (GE), pp. 5-48.
2.
Pilgrim: Reactor Theory (GE), pp. 7-18 through 7-20.
. KA 29008K120 (3.3/3.4) 292OO8K121 (2.9/3.0) 293OOSK105 (2.7/2.8) . t e 4------..m,,--, , - - - ,3m., ---,,-p . -, ,_---mv-.,-r__,- ,w,,, ,,,--,---,w-%*, ,,,. ,,, - f -
+ww--
,,,
.. _ _ . . - - _ - . - - _ . , . , 111. PRINCIPLES OF1 NUCLEAR POWER' PLANT OPERATION.
PAGE
. - THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW .. ANSWERS -- PILGRIM-87/12/08-MOON.
D.
' ANSWER' 1.05 (3.00) , PO exp (t/T) Ehere T = ceriod, t = 180 seconds, a.
P1 = + P1/PO = 10 C+0.5) T = t/LN(10) T = 78.2 seconds [+0.5] ' (25/15)*100 = 167 [+0.5] Then t' = T LN(P2/P1) where P2/P1 = (78.2) LN(167) 't' = (78.2) (5.12) = 400 seconds (or 6.7 minutes) [+0.53 = p r-pw spijt/ b.
The reactor. c r i ;4 increase is governed by how cuickly the neutron population can increase.
The same holds true on a power decrease; however, for a power decrease the neutron population i s dominated by the. longest lived delayed neutron prec.ursor C+1.03.
(This decays with -80 second period.)
REFERENCE.
1.
Pilgrim Reactor Theory, p.
3-15.
2.
Pilgrim Systems Reference Texts, IRM, Figure 8.
KA 292OO3K108 (2.7/2.8) 292OO3K106 (3.7/3.7) . ANSWER 1.06-(1.50) . a.
seven to eleven hours C+0.753 , !-'b.
seventy hours +/- ten C+0.75] REFERENCE 1.
Pilgrim: Reactor Theory, p.
6-11.
i KA 292OO6K107 (3.2/3.2) l
. . , y v.
c.
.
,+,,-e-w
.--,.,,,,--#w.~,w.--.rn-wow,r--rem-r--+.w,-,_m,r,-,3..n.,w%-,m=-,-73wwi-.,, ---*---,-v-e , - -, g ,ce,
.. -. .. - - _ . - - , , - . l ~li__EBINCIELES_gE_NYCLE@B_EgyE6_EL9NI_ GEE 8@llgN PAGE
- t.
THERMODYNAMICS, HEAT TF4NSFER AND FLUID FLOW . TANSWERS -- PILGRIM-87/12/OB-MOON.'D.
. m . ANSWER-1.07 ( 3. 00)- , a.
1The fuel claddino temo'erature'can exceed 2200 decrees F durino ' a DBA LOCA C+1.O] OR gross cladding failure C+1.O]. (Either is correct for full credit) .b.
1.- Increase'C+0.5] due to increase in inlet subcooling C+0.53.
2.
Decrease C+0.53 due to decrease in core t' l o w C+0.5]. REFERENCE 1.
Pilgrim: HT&FF (GE), pp. 9-19 througn 9-24.
~ KA 293OO9K111 (2.8) 293OO9K123 (2.8) 293OO9K122 (2.9) ANSWER 1.08 (1.00) The increase in void generation due to the power increase down low in a bundle (next to the control rod withdrawn) will result in a , negative reactivity feedback throughout the length of the bundle.
In this case, the positive reactivity due to control rod withdrawal was less than the negative reactivity due to local generation of voids C+1.03.
REFERENCE 1.
Pilgrim: Reactor Theory, pp.
4-16, 4-17, and 7-13.
2.
Station Nuclear Engineering (GE), Vol.
1A, p.
4-2.
i 104 292OO5K104 (3.5/3.5) 292OOOK119 (3.1/3.2) . . --y .eq wr-- -,-g---4 .n-mc.,---.--m-q. m ggm, w-m-om~- ,v--m.m,, w e gm. g g- .m .
___ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _, _ - - _ - - _ - - _ __;_.
_ __ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - - - _ - _ _ _. - _ - _ _ - _ - _ - _ - -. - si h__EBI EIELES OF NUCLEAR POWER' PLANT OPERATION.
'PAGE .24 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ..
- ANSWERS -- PILGRIM ~
-87/12/08-MOON.
D.
ANSWER,. 1.09 (2;00) , 1.
pumos are l'o c a t e d ~be16w normal water level 2.
minimum speed interlock i s enforced when feedwater flow is less than 20 oercent 3.
at high power operation, feedwater subcooling provides increased net positive suction head
recirculation pump runback on low reactor vessel water level ' 5.
recirculation pumo. trips (or won't start) if suction valve is'not full open (90 percent) 6.
recirculation pumo runback (speed limit) if discharge valve-is not full open (90 percent) 7.
recirculation pump trio on low-low water level 'O.
recirculation cump trip i f discharge valve control switch is p1 aced in CLOSE Any four (4) (+0.53 each, +2.0 maximum.
, REFERENCE 1. - Pilgrim Studert Guide 0-RO-02-06-02. pp. LG-3 and LG-4 . 2.
Pilgrim: Systems Reference Texts. Reactor Recirculation, p p.- 21,.22, and 35.
. KA 202OO1K402 (3.1/3.2) 291004K106 (3.3/3.3) 293OO6K110 (2.7/2.8) e. m d. A c / a ; x 4 / m M n /0 %A ' a 7.
wJa agm,,a.v a ' ' dfc kk.
a pp 10 L GL-u g i,J h a,As.,anoca & h 1 L~i.
, 4N+' ' . 'V/f7 . . . r -......., ,-.. ._ -. - ,_.,_...-.,_,._,,,,_,-m_~ . _.. - - _ ~. _ _. _ _ _,.. -. ,.. ~ -.,_ _,--,,_ -, -
. Li__ESINCIELES_gE_ NUCLE @@_EgWEE_EL@NI_QEEE@llgN.
PAGE
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW .. ANSWERS -- PILGRIM-87/12/08-M00N.
D.
ANSWER 1.10 (3.00) , c.
Head increases C+0.53.' The pumo is still outting the same amount of v.or k into the fluid. therefore the same delta cressure across the pumo, so as the suction pressure increases so will the discharge head C+0.53.
b.
Head increases C+0.53.
As system r esi st anc e to flow increases, cumo head increases C+0.53.
c.
Head decreases C+O.53.
As temperature increases. system resistance to flow decreases (lower viscosity). therefore heao decreases C+0.53.
REFERENCE 1.
Pilgrim: HT&FF (GE). pp. 6-95 throuah 6-102. and 6-10c.
2.
Pilgrim: Student Gu'i d e. HT&FF, po. 6-lO9a and b.
KA 291004K113 (2.6) ANSWER 1.11 (2.00) a.
(Although turbine efficiency would increase) overall plant efficiency would decrease C+0.251 because the heat rejected to the circulat'ing water must be added to the feedwater by the reactor C+0.753.
b.
Air ejectors would still be needed in order to maintain condenser vacuum because air i n '. e a k a g e i s not the only source of noncondensibles to the main condenser C+0.53.
Other NC H2(and fission product gases)C+0.53.
j gy, include radiolytic 02 and r M REFERENCE 1.
Pilgrim: HT&FF (GE), pp. 5-12 through 5-14 KA 293OO4K112 (2.9/3.1) 291006K118 (2.8/2.9) . .
. 2i__8(@NI_QE@lGN_lyC(yQ1N@_@@ Eely _@NQ_EME6@ENCY_SYSIEMS PAGE
ANSWERS -- PILGRIM-87/12/08-MOON.
D.
- ANSWER 2.01 (2,00) l+1.03
k QWuhp ).l(pg ' a, local fuel damage (bv generating a rod withdrawal Dlock) kiew b.
units = volts [+0.53.
number of operable LPRM inouts can be , f'/fk7 calculated (bv using 1 volt oer coerable inout) [+0.53 REFERENCE 1.
Pilgrim: Systems Reference Texts. RBM. op.
I and 11.
KA 215002SGO4 t3.3/3.4) 21500bK102 (3.2/3.1) 2150024402 (2.9/2.9) ANSWER 2.02 (3.00) a.
yes [+1.03 b.
no [+0.53: must insert the rod [+0.53 OA J/gf( r.. no [+0.53: must raise the grapple fully,gt move refuel olatform away from core [+0.53 , / /j7 REFERENCE 1.
Pilgrim: Systems Reference Texts, Refueling, pp.
7, 28, and 29.
KA 234000K502 (3.1/3.7) 234000A302 (3.1/3.7) l l l l l . .
._ - -.. - . 2:__EL@NI_Dg@]QN_JUCLUp]Ng_58EEIy_@Np_EUER@ENCy_SygIEU@ PAGE_ 27 l ANSWERS -- PILGRIM - - -87/12/08-MOON.
D.
. ANSWER 2.03 ( 3. 00).
, . -a.
2.
f.
2.
b.
1.
g.
2.
c.
1.
h.
2.
d.
1.
i.
1.
e.
4.
j.
2.
C+0.33 each REFERENCE 1.
Pi1 grim: Procedure 2.2.46.
p.
5.
2.
Pilgrim: Systems Reference Texts. TBCCW, pp.
I and 2.
3.
Pilgrim: Systems Reference Texts. SSW.
p.
1.
Pilgrim: Systems Reference Texts. RBCCW, p.
21.
KA 203OOOK116 (3.1/3.2) 259001K605 (2.7/2.7) ANSWER 2.04 (2.50) 4.
1.
(high level switch provides) a start permi ssi ve to one sump pump (when pump is in AUTO mode) 2.
(high high level switch provides) a start permissive for / U .NW second sump pump _ _. % 3 m.m.
,s....- " ^ f '__'.. m& " _ " _ [~ '" .Ml, ,I'7 ^ "~ ' _ m.,_m _ a . m ., -,um.a.. . . - . t#* 7I . C+D<&3 each st}/3}87 b.
Place the H2/02 valve control switches in CLOSE C+0.333 place the override keylock switches in OVERRIDE C+0.343 and place the H2/02 valve control switches back in OPEN C+0.333.
REFERENCE 1.
Pilgrim: Systems Reference Texts. DLD, pp. 3 and 13.
KA 223OO1K104 (3.2/3.3) 223OO1K110 (3.4/3.6) . . . . . - . . . . . . . . . .
l " 2i__E(@UI_gESigy_lMG6gplNQ_g@EEIX_999_EUg8GEUGy_SYSIEUS PAGE
-87/12/08-MOON.
D.
ANSWERS -- PILGRIM - ANSWER 2.05 (2.00) a.
1.
low accumulator nitrogen pressure C+0.53 (l ess' than or eaual to 950 psig7 2.
high accumulator water l evel C+0.53 (greater than or equal to 37 cc leakage cast the accumulator seal) c.
1.
inlet scram C+0.33] 2.
outlet scram C+0.333 3.
open C+0.343 (Items 1.
and 2.
can be in either order.)
REFERENCE 1.
Pilgrim: Systems Reference Texts. Reactor Manual Control, p.
3.
KA 2OlOO2 GOB (3.6/3.4) 201002G11 (3.4/4.0) 201002G12 (3.4/3.2) 201002K101 (3.2/3.2) . .
._ . . .. . ~ . . 2i__E(ANI_ DESIGN _lNC(yD{NG_59EEIY_gND_EMEEGENCY_SYSIEMS.
PAGE 29: ANSWERS -- PILGRIM-87/12/08-M00N.
D.
- ~ ANSWER 2.06 (3.00)- c a '. decrease as the SJAE supolv valves close (or becaus'e offgas isolates) ' b.- decrease because RBCCW HX. bypass valve closes, providing maximum cooling to RBCCW c.
decreases as CRD FCV's close ~d.
. decreases as FW regulating valves fail as-is and condensate oumo and FW pumo recirc valves fail open e.
decreases because bleeder trio valves fail shut und soill valves fail coen) thus eliminating all FW heating f.
tank level indication goes to zero since level instrument' recuires air to operate (but no change in actual tank level) .
C+0.53 each REFERENCE-1.
Pil gri ms - Systems Student Guide, Instrument Air, pp.
12 and 13.
, KA 295019AKO1 (3.8/3.9) 295019AK102 (2.9/3.0) 295019AK103 (3.2/3.3) 295019AKIO5 (3.4/3.4) 295019AK106 (2.8/2.9) 295019AK115 (2.3/2.6)
! I i ANSWER 2.C7 (2.50) N a.
Normal - RPS MG sets C+0.53 (or MCC B-23 and MCC B-22 , Alternate - 480V MCC B-10 C+0.53 ' i IAAt A El - '-- ' - ' ' - '=' -- --* **i^^ ~^~^--) b
P h-gas system radi ation , (oh monitors) IL)b [ 2 l . 6.
(main steam line radiation monitors) C+0.53 each, C-0.53 for incorrect answers REFERENCE ! ' 1.
Pilgrim: Systems Reference Text, Reactor Protection System.
, pp. 9-3 and 36-2.
' KA 212OOOA202 (3.7/3.9) 212OO2K201 (3.2/3.3) , ,- r - - - - - -= '-we- --w-*
e-e
-+--,,---.----w- ~,~~.----w r--e.-<- +--e---- - - - - + -* w e-
. -. 2 __E68NI_QESIGN_INC(yDJNG_@@EEIY_9ND_EMESGENCZ_5YSIEUS PAGE
ANSWERS -- PILGRIM-87/12/08-MOON.
D.
- ' ANSWER 2.08 (3.00) a.
Will continue to operate C+0.53.
Operation under t'h e s e conditions would allow contamination of the RCIC room and atmosphere from turbine and valve steam leakage C+0.53.
StLC .U i5de$t. e l u c c w 1]I C Ni%rt?
- jffg Willfhontinuetooperate C+0.53.
Nc ac'/e se cI3ect c r.
7% b.
n/jpg7 cp;r 2 + 4 mm C+0.53.
4Isolouton is at a7 p s i g. ' ', W2fSNv2 Wh c.
RCIC will nop/contihue to coerate C+0.53.
RCIC turbine will trip due to,,.i..mfii-iact val art ur t o - hol-d r = d - c nhel k,(I A/[/9 97 , c U., st L i c-val ve opens C +0. 5 3.
REFERENCE 1.
Pilgrim: Systems Reference Texts. RCIC. pp.
3.
5.
and 19, Figure 6.
KA 217000K405 (3.2/3.5) 217000A207 (3.1/3.1) 217000AIO4 (3.6/3.6) 217000GO7 (3.8/3.7) ANSWER 2.09 (1.50) a.
No C+0.53.
(If a core spray line breaks inside the shroud, the differential pressure indicating switch will detect reactor pressure inside the shroud as usual: therefore, no abnormal differential pressure will be indicated.)
b.
The core spray loop can perform a flooding function but its spray will not provide full core spray coverage.C+1.03.
REFERENCE 1.
Pilgrim: Systems Reference Texts, Core Spray, p.
10.
KA 20900lKil3 (2.8/3.0) 20900lK404 (3.0/3.2) > l - . - I . . . . -. . . .. . .... - . - . . . - . . .
. 7 __E60NI_QEglGN_lNC(gylNg_g@ Eely _@Ng_EdE5GENCy_SYSIEdg PAGE
ANSWERS -- PILGRIM - -87/12/08-MOON.
D.
ANSWER 2.10 (2.50) . a.
1.
ref uel floor E+0.5] 2.
drywell C+0.53 - 3., suppression pool C+0.5] ).2 (""1 0'Eb kJ/lIG b.
The drvwell isolation c;ct button-must first be r eset C+1.03.
p
REFERENCE 1.
Pi l gri m: Procedure 2.2.50.
p.
10.
2.
Pilgrim: Procedure 2.4.147 p.
3.
3.
Pi l gr i m: Systems Reference Texts. SGTS. pp. 3 and 12 and Figure 1.
KA 261000K101 (3.4/3.6) 261000K102 (3.2/3.4) 261000K103 (2.9/3.1) 261000K401 (3.7/3.8) . ! l . .
- . _.
' . 4 __ INS 16UMENIS_AND_CgN16gLS PAGE
ANSWERS -- PILGRIM-87/12/08-MOON.
D.
- b OO[8 ANSWER 3.01 (3.00) , a.
Bvpassed when the mode switch is NOT in RUN C+O.753.
n _
'b.
Auto bypassed after-(2 sec.) time del av C+0.753.
c.' Auto bypassed if reactor power (45 percent (as indicated by turbine first' stage pressure of 305 pstg) +0.753.
ght d.
Manual bvpass switchh44 in BYPASS with mode switch in j $7 SHUTDOWN or REFUEL C+0.753.
REFERENCE 1.
Pilgram: Systems Reference Texts. RPS. pp.
17 through 25.
KA 212OOOK412 (3.9/4.1) 212OO4K408 (4.2/4.2) ANSWER 3.02 (1.50)' 1.
an SRM HI-HI will no longer initiate a full scram C+0.753 2.
the RPS will no longer trip (full scram) on anv ONE nuclear instrumentation trip (noncoincidence) (i.e.. RPS will be in the coincidence mode) C+0.753 REFERENCE 1.
Pilgrim: Systems Reference Texts. RPS, pp.
12 and 14 . KA 212OOOK101 (3.7/3.9) 215004A303 (3.6/3.5) 212OOK411 (3.3/3.5) . . (
. ., Il__INglgydENI@,@ND_CONIBO(g PAGE
, ANSWERS.-- PILGRIM - -87/12/08-MOON, D.
-
- ANSWER 3.03 (2.50) a.- ADS valves remain as is C+0.53 ' b.
. ADS valves close C+0.~53 c.
ADS valves close C+0.53 l d.
ADS valves ooen C+0.53
- e..
ADS valves open C+0.53 REFERENCE 1.
Pilgrim: Student System Guide. ADS. pp. 4 through 10.
2.' Pilgrim: Systems Reference Texts. 405.
pp.
3 throuch 20.
KA 218000K501 (3.3/3.8) 218000K404 (3.5/3.6) 218000A205 (3.4/3.6) 218000K602 (4.1/4.1) 218000A105 (4.1/4.1) , ANSWER 3.04 (2.00) kM2.c.LL.
a.
l.
5 c-c ;: C+0.53 , h(m iebu& ispyd Y*\\S M#lW lY/9lA7 2.
decrease / C +O 53 . ~% noc _ b.
narrow range C+0.53 , jy'/.t f /g/)f ? / c.. fuel zoney C+0.53 D#"#
- T REFERENCE
, 1.
Pilgrim: Systems Student Guide Non-Nuclear Instrumentation, pp. 23 through 25.
2.
Pilgrim: Systems Reference Texts, Nuclear Boi l er Instrumentation. pp.
5, 6.
and 22.
KA 216000K501 (3.1/3.2) 216000K507 (3.6/3.8) ' 216000K510 (3.1/3.3) 291002K108 (2.8/2.9) 216000K101 (3.9/4.1) . I r
. e . . . . . . .. . .. .. .
. ._ _.. . _ .m ... . . - 2c__ INS 18(MENIS_8ND_CQN18QLS PAGE
' ANSWERS -- PILGRIM-87/12/08-MOON.
D.
- . ANSWER 3.05 (2.00) t c9 -{b unvttka h5h p j7 ROD BLOCK - due to flow comparator INOP 3 C+0.53 - due to flow biased trip (70 > (0.58W) + SO: W=O) C+0.53 "' - RBM f low bi ased trip C+0.53 ' .ONE-HALF SCRAM Ch. A - due to ficw ' biased trit (70 > (0.58W) + 62: W=O) '. 53 REFERENCE 1.
Pilgrim: Systems Reference Texts. APRM. po.
4, 5.
8,
and 19.
, }O4 215005K116 (3.3/3.4) 215005K607 (3.2/3.3) 215002K101 (2.9/3.0) 212OOOK603 (3.5/3.7) , , t ' ANSWER 3.06 (3.00) , [n4 7() , a.
Adju ts the gain of tt.e RBM channel - 'U.731 to compensate , [[/ I O.'5'v-OR bypassed LPRM's / for variations in local power - C+0.753.
/
, b.
1.
Joystick C+0.53 2.
edge rod selected C+0.53 ' 3.
reference APRM < 30 percent C+0.53 REFERENCE
1.
Pilgrim: Systems Reference Texts, RBM. pp. ~, 6, 8, 15, and 18.
-KA 215005K502 (2.4/2.5) 215005K403 (2.9/3.0) , 215005A403 (2.8/2.8) a
' L . l . . I I - .. - .. ..-..- -.. -.. ... -. -.- - - -...-... -.- -.- -. - . ...
, ' . j < ~ ~ 5'E__16@l5UMENI@_AND_COU150L@ PAGE
ANSWRS -- P ILGRIM-87/12/08-MOON.
D.
- -
ANSWER.
3.07 (3.00) c.
Detectors are located in the steam pipe tunnel, n e x't to main' hSIV's)C+1.03.
{ steampiping.(downstr6amof the outboard kP'
I b.
1.
reactor scram [ S. a f 2.
mechanical vacuum pumo isolation (trip) l*0_"' . 3.
main team line drains close J'O.51 MO4l # M */ 3 bb b **" * ao %h fertt V*I'(1,1 ?{ y 4+ PCor g c.
4v = is initiated '1 3 c;rr r. e ts: c.s offgas hold-up $h'l cimer u uoi
volume isolatesy C+0.53.
l 6b4.0 (13. p'iew hitwb k't N O Me-l REFERENCE 1.
Pilgrim: Systems Reference Texts. PRM. pp. 6 cnd 22.
KA 272OOOK101 (3.6/3.8) 272OOOK102 '3.2/3.5) 272OOOK402 (3.7/4.1) ANSWER 3.08 (2.00) .
a.
Reactor pressure is decreasing (given) due to EPR ca sing the g[k control valves to open C+0.333.
BPV's r 2i-cirred 3C+0.333.
jy/gfjy Reactor pressure continues to decrease until MSIV's isolate on low steam line pressure (with the mode switch in RUN) C+0.343.
b.
On power loss to EPR. EPR closes the control valves which causes reactor pressure to increase C+0.333.
When reactor pressure reaches the MPR setpoint. the MPR will control the control valves to maintain oressure C+O.343.
BPV's do not open C+0.333.
! REFERENCE I ! 1.
Pilgrim: Systems Student Guide. Main T.irbine Steam System.
pp.
16 through 19.
KA 24100K419 (3.6/3.7) 24100A102 (4.1/3.9) 24100A107 (3.0/3.7) 24100A308 (3.8/3.8) l ! . . ( l L
. 11__INSI@yMSNIS_gND_CONISOLS PAGE
ANSWERS -- PILGRIM - -87/12/08-MOON.
D.
ANSWER 3.09 (3.00) a.
Both recirculation oumos run back to 65 percent s o e'e d C+0.53 due to the automatic r'unback interlock with soeed limiter (#2) C+O.5]. b.
Recirculation cump A runs back to 28 percent C+0.5] due to the discharge valve not full ooen interlock with speed limiter (#1) (+0.25]. Recirculation pumo B speed will be unaffected C+0.25]. c.
Both pumps will runback to 28 percent C+0.53 due to the FW flow signal (l ess than 20 cercent) interlock with speed limiter (#1) C+0.53.
REFERENCE 1.
Pilgrim: Systems Student Guide. Reactor Recirculation, pp.
and 10.
2.
Pi l gri m: Systems Reference Texts. Recirculation Flow Control, pp.
2, 5, 6, and 12.
KA 202OO2K305 (3.2/3.3) 202OO2K604 (3.5/3.5) . .
. . . ... . .- - ., - - .. . - -.- , . Ti_,]N@I69M@yl@,9Np,CgNI@QLQ' PAGE
, ANSWERS - PILGRIM-87/12/08-MOON.
D.
- ANSWER 3.10 '( 3. 00 )
a.
FW Flow Level increases C+0.253;due'to large flow' error signal causing FRV to'coen C+0.253.
. Probable turbine. trip on-high level C+0.253.
b.. STM Flow: Level decreases C+0.253 due to large flow error signal causing the FRV to close C+0.253.
Low level scram will occur C+0.253.
c.
'RX Water Level: Level increases C+0.253 due to large level error. signal causing FRV to open fully C+0.253.
Turbine trio on hi -l evel will occur C+0.253.
d.
Density compensation: Level increases C+0.253 due to the indicated level being; lower than actual level causing an error . between level and FW-flow and opening the FRV-C+0.253.
No ~ turbine trip or scram will occur C+0.253.
- REFERENCE 1.
Pilgrim Systems Student Guides. Feedwater System, pp. 44 through 48.
I 2.
Pilgrim Systems Reference Texts. FWLC. pp.
19 through 22.
KA 259002K603 (3.1/3.1) 259002K604 (3.1/3.1)
259002K605 (3.5/3.5) 259002K606 (2.6/2.7) i'
1 l [ h a l I . i . i i . _,.,... . _ _.. _ - - _ _ _,. _. -..... _ _.. _, _ _. . _ _. _ _, - _ _ -, _.. _ _ _. _ -. _,.,. _... ,. -. - . _ _. -.
_ _ _ _ - . . ft__BBOCEggBE5_;_UOBb863,@BNOBU@61_EdgBGENgy_ANQ PAGE
69919L991C9L,CgNJggL .. ANSWERS -- PILGRIM-87/12/08-MOON.
D.
ANSWER 4.01 (2.00) , 1.
excessive sumo pump operation due to increased leakage to - drywell equipment drain sumo or drywell floor drain sumn 2.
abruot increase in drywell humidity (Panel CBS) 3.
significant drvwell pressure increase 4.
significant drywell temperature increase C+0.5] each REFERENCE 1.
Pilgrim: Procedure 2.4.14 Leaks Inside the Primarv Containment, p.
2.
KA 295010AK103 (3.2/3.4) 295010AA106 (3.3/3.5) 295010AA202 (3.8/3.9) 295010AA204 (2.8/3.0) ANSWER 4.02 (3.00) a.
1.
recirculation system delta pressure deviatior. (as d et er mi n ed by test) 2.
sudden change in core delta pressure without accompanying change in recirc pump speed 3.
unexpl ained changes in recirculation flow Any two (2) C+0.753 each.
+1.5 maximum.
b.
1.
monitor alarms and instrumentation to determine type of system malfunction that has occurred C+0.753 2.
if it appears that a jet pump probl em ex i st s, refer to the Technical Specification LCO C+0.753 REFERENCE 1.
Pilgrim Procedure 2.4.23, p.
2.
KA 202OO1K601 (3.5/3.7)
, . I . . .. . ... .
. 9 t_ _ESQGE QQBE 5_; _dgBd@6 t _980060@6 t _ s DE EG E ggy,@yD PAGE
6891969 GIG 86_GQNI@QL .. ANSWERS -- PILGRIM-87/12/08-MOON.
D.
ANSWER 4.03 (3.00) , a.
1.
announce the event over the station paging system 2.
reduce recirculation pump speed to minimum 3.
scram the reactor by placing the mode switch in shutdown 4.
verify all control rods have fully inserted 5.
after 20 seconds trip the main turbine 6.
verifv the cenerator output breakers are open
trip the reactor feed pumps 8.
if necessarv. evacuate the MCR C+0.3] each b.
1.
SRV's 2.
HPCI 3.
RCIC Other answers may be acceptable.
C+0.23 each.
+0.6 maximum.
REFERENCE 1.
Pilgrim Procedure 2.4.143, pp. 2 and 3.
KA 295016SG10 (3.8/3.6) ANSWER 4.04 (1.50) ! l Must confirm by at least two C+0.5] independent C+0.5] indications that the initiation was inadvertent C+O.5]. REFERENCE ,
I 1.
Pilgrim: Procedure 2.4.35, p.
2.
. KA 209001G01 (3.9/4.1)
, .__ _ _
_ < . . 9E__EBOrjQQBES_;_NOBd@bt_@BNOBd@(t_EdEBQ[NQy_@NQ PAGE
Be91969G1Ge6_GQNIGQ6 .. ANSWERS -- PILGRIM-87/12/08-MOON.
D.
' ANSWER 4.05 (3.00) . a.
allowable exposure = 5(23-18) 25 R - 23 R = 2 R C+1.03 = b.
allowable exposure i s IOOO' mrem /atr C+0.53 remaining exposure = 1000 - 250'= 750 mrem C+0.53 750 mrem 3.75 hours C+0.53 stay time =
=
200 mrem /hr c.- 2ero mrem the cannot. receive any occupational exposure at PNPS) C+0.53 ' REFERENCE 1.
10CFR20.101.
2.
Pilgrim PNPS Radiation Exposure Control Program. 6.2-001, pp. 2.through 5.
KA 294001KlO3 (3.3/3.8) ANSWER 4.06 (2.00) .- 1.
monitor alarms and instrumentation to determine the type of system malfunction that has occurred (determine direction of rod travel) C+0.5J - 2.
select the drifting control rod (s) and apply a withdr.a or insert signal to stop the rod (s) C+0.753 3.
if the two rods are in a nine rod array, and drifting in, scram the reactor C+O.753 REFERENCE 1.
Pilgrim: Procedure 2.4.11, p.
2.
KA 295014AA104 (3.2/3.3) 295014AA105 (3.9/3.9) 295014G10 (4.0/3.9) . .
. 31__8399E990EE_2_U98U661 9Ed90Ueb _{[[6g[ygy_AND PAGE 41-Se9196991Geb_GQNigg( .. . ANSWERS -- PILGRIM-87/12/08-MOON.
D.
ANSWER 4.07 (2.00) , c.
1.
- b.
OR 5.
(EITHER is acceptable for full credit) c.
5.
. d.
2.
C+0.53 each REFERENCE 1.
Pilgrim: Tagging Procedure 1.4.5.
op. 7 and 8.
KA 294001K102 (3.9/4.5) ANSWER 4.08 (3.00) a.
1.
drvwell temperature above 152 degrees F 2.
drywell pressure greater than 2.5 psig 3.
torus water temperature above 80 degrees F
torus water level not between 127 and 130 inches (-3 to-6 on narrow range) 5.
primary containment hydrogen concentration above 4 percent . C+0.53 each b.
of f site radioactive release rate above the of f site release rate which requires an alert C+O.53 REFERENCE 1.
Pilgrim: EOP-03, Flowchart.
2.
Pilgrim EOP-05 Flowchart.
KA 295024SG11 (4.3/4.5) . e
!- ,, -
- j ;
- E.
- PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE
BOD 196gGig@6_ggNIBQ6 .. ANSWERS -- PILGRIM.
-87/12/08-MOON.
D.
ANSWER-4.09 (1.50) , a.
Secondary Containment ~ Control (EOP-04) b.
-RPV Control (EOP-01) -c.
Secondary Containment Control (EOP-04) C+0.53 each REFERENCE 1.
Pilgrim: EOP-04. Flowchart.
2.
Pilorim: EOP-01. Flowchart.
KA 295036G11 (3.8/4.1) 295025G11 (4.2/4.3) 295034 (4.2/4.3) ANSWER 4.10 (1.50) a.
145 degrees F b.
50 degrees F c.
100 degrees F/hr d. So 1HIN> p si g Qvw syl8f87 e.
145 degrees F f.
5 percent C+0.25] each REFERENCE - 1.
Pilgrim: Procedure 2.1.9 pg.
A-2.
2.
Pilgrim: Technical Specifications. Section 3.6.A.
3.
Pilgrim: Procedure 2.2.92, Precaution A, p.
7.
4.
Pilgrim Procedure 2.2.93, Precaution D, p.
7.
5.
Pilgrim: Procedure 2.1.3, Li mi tati ons A. 3, B.2, pp. 2 and 3.
. .
e 9 __ES99E99855_ _SgSd863_9pygSd963_gdgEgghgy_99D PAGE
689196991996_G90I596 . ANSWERS -- PILGRIM-87/12/08-MOON.
D.
ANSWER 4.11 (2.50) . 6.
to minimize feedwater' nozzle thermal dutv [+1.03 b.
to ensure a natural circulation flowcath exists [+1.5] REFERENCE 1.
Pilgrim: Procedure 2.1.3.
p.
2.
Pilarim: Procedure 2.1.5c.
n.
3.
KA 239001K317 (3.2/3.3) 202001K507 (3.3/3.4) l l l l l l l l i
e .
e - A +La ( f,,,, (, } 2, o U. S. NUCLEAR REGULATORY COMMISSION SENIOR AEACTOR OPERATOR LICENSE EXAMINATION , O FACILITY: _PI(981M__,______________ REACTOR TYPE: _@WR-@E] ________________ DATE ADMINISTERED _@Zfl?!9)________________ E X AM,1 NER: _@RATTQN_C)_6pg_q._6.__
3 CANDIDATE: _____ M _k $_~f~[ h ______ INS 180CI]QN)_Ig_CONp]p6IE1 Usa separate paper for the answers.
Write answers on cne side only.
Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question.
The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
X OF lATEGORY % OF CANDIDATE'S CATEGORY . _VALUE__ ______________CATEQORY_____________ VALUE TOTAL SCORE L2Et99_. 2E:99 ________ 5.
THEORY OF NUCLEAR POWER PLANT ___________ OPERATION, FLUIDS, AND THERMODYNAMICS 3! 99_- _2E199 ________ 6.
PLANT SYSTEMS DESIGN, CONTROL,
___________ AND INSTRUMENTATION ,2Ez99._ 3E 99 ___________ ________ 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL .25299__ _3E299 ________ 8.
ADMINISTRATIVE PROCEDURES, ___________ CONDITIONS, AND LIMITATIONS 199199 _ ________% Totals ___________ Final Grade
l All work done on this examination is my own.
I have neither given ncr received aid.
___________________________________ Candidate's Signature l i i l __ __ _ _-
~ . . NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS ' 1 ring 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.
Restroom trips are to be limited and only one candidate at a time may-leave.
You must avoid all contacts with anyone outside the examination roca to avoid even the appearance or possibility of cheating.
Use black ink or dark pencil gnly to facilitate legible reproductions.
- Print your name in the blank provided on the cover sheet of the exacination.
< Fill in the date on the cover sheet of the examination (if necessary).
Use only the paper provided for answers.
< Print your name in the upper right-hand corner of the first page 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 gnly gn gge side of the paper, and write 'Last Page" on the last answer sheet.
Nu:ber each answer as to category and number, for example, 1.4, 6.3.
. Skip at least three lines between each answer.
1. Separate answer sheets froa pad and place finished answer sheets face do:n on your desk or table.
. Use abbreviations only if they are coanonly used in facility litetatute.
The point value for each question is indicated in parentheses after the . qu3stion and can be used as a guide for the depth of answer requirud.
D. Shou all calculations, methods, or assumptions used to obtain an answer to mathematical probleas whether indicated in the question or not.
f. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
l If parts of the examination are not clear as to intent, ask questions of . l the examinet only.
, f.Youcustsignthestatementonthecoversheetthatindicatesthatthe j tork is your own and you have not received or been given assistance in l cocpleting the examination.
This must be cone after the examination has l been completed.
l l ,
- . . ).'When you ccaplete your examination, you shall: a.- Assemble your examination as follows: (1) Exam questions on top.
(2) Exas 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 all scrap paper and the balance of the paper that you did not use for answering the questions.
d.
Leave the examination area, as defined by the examiner.
If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
I l l l L.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
. 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLU]DS _AND PAGE
3
..IUERMgD]NAgjCS QUESTION 5.01 (3.00) -Following an automatic initiation of LPCI at a reactor pressure of 350 psig, reactor pressure decreases to 200 psig.
For each of the parameters listed below, determine any . change that may occur (i.e. increase, decrease, or remain the same).
BRIEFLY EXPLAIN why the parameter-changes or remains the sace, a.- LPCI injection flow (1.03 b.
LPCI pump discharge head ( assume constant NPSH ) (1.03 c.
LPCI pump' speed (RPM's) (1.03 QUESTION 5.02 (2.50) Regarding core thermal limitst a. The process computer output, CMFLPD, is used to sanitor WHICH core thermal limit? [0.53 ' b. WHICH core thermal limit ensures peak cladding traperature will not exceed 2200 deg.F following a LOC 47 (0.53 c. What is the failure mechanisia and the limiting condition for LH6R7 (1.03 d. What potential problems and adverse effects would occur in a f uel bundle if it were operated in excess of the MCPR limit? [0.53 ,
(***** CATE60RY 05 CONTINUED ON NEXT PAGE *****) t
. . 31..IBEg81,QE 89C(E63,&gy!B,EL6BI,9fE8611903,[LylpS AyQ PAGE
3 ,IUE859pyNSD]CS QUESTION 5.03 (2.50) a. Consider two control rods. Both rods are at notch position 16.
Rod A is located near the center of the core and rod B is located at the core edge. The reactor scraas after operating at high power for a long time. A hot startup was performed and power reached 30% about ten hours after the scram. To add the LEAST reactivity at this time with a one notch withdrawl, WHICH rod (A orb) would you choose and WHY? [1.5) b. Shaping control rods are (DEEP, SHALLOW,1NTERMEDIATE) rods that are used to change the power profile because they (ARE, ARE NOT) af f ected by shadowing.(Choose one answer in each parenthesis) (1.0 QUESTION 5.04 (2.00) Indicate whether the following statements concerning fission product poison behavior are TRUE or FALSE for your reactor.
IF FALSE, change the statement so that it is correct.
(Consider each statement separately) a) Equilibriua xenon concentration at 50% power is approximately half of the equilibrius concentration at 100% power (0.5) b) Equilibrius sanarium concentration is the same for all power levels.
(0,5) c) Both xenon and samarium concentrations increar,e feardiately after a reactor shutdown from high powers.
(assume equilibrius had been reached before shutdown) (0.5) d) Xenon-135 decays with a half-life of about 4.5 hours while samarium-149 is stable.
[0.5) i ! QUESTION 5.05 (1.50)
The subcooling of condensate exiting the condenser is termed condensate depression.
a) Briefly explain why condensate depression is necessary? [1.0 l b) Briefly explain why excessive condensate depression will shorten the fuel cycle.
(0.5 (***** CATEGr.Ry 05 CONTINUED ON NEXT PAGE *****)
. . !Qs. 10EQ81-QE.Nyq(E@B,&QWE8,E(@NI,Q&[8@l[QN,[(y[Q$5,@ND PAGE
1 , IME85991NQQ[QS QUESTION 5.06 (3.00) Will the following cause core reactivity to increase or decrease? Briefly explain for each why reactivity will change.
Assume reactor is at full power.
[1.03 a) Loss of a feedwater heater - b) Sudden increase in reactor pressure (prior to a reactor scran) (1.03 c) Build-up of corrosion products on the fuel pins.
(1,03 . QUESTION 5.07 (2.00) a.
Which one of the following has NO effect on the available net positive suction head (NPSH) of a recirculation pump? (1.03 1. RPV level in the downconer region 2. The temperature of the feedwater entering the vessel 3. Reactor steam done pressure 4. The number of operating rectrulation pumps b. A recirculatiov pump operating at a constant speed draws least current (Choose one) (1.03 1. At runout 2. At operating conditions 3. At shutoff head (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
. . 5 _.IBE981.9E_NUQ([@B,[9MEB [(@NI_QEE8911QN _[(Ujpj3,6ND PAGE
1
. ISEBD9915051gs QUESTION 5.08 (1.50) Which of the f ollowing is the correct value f or a cooldown rate, given an initial reactor pressure of 965 psig and a reactor pressure of 515 psig i hour later ? [1.53 a. 50 F/hr . b. 70 F/hr c. 90 F/hr 4.
110 F/hr , QUESTION 5.09 (2.50) You are performing a reactor startup and the reactor is critical at a power level of 2 kW. The reactor operator withdraws a control rod 2 notches and reactor poser increases with a doubling time of 90 seconds.
Ansser the following showing all works a. "hat is the reactor period calculatet.n seconds (0.753 b. Assuming no further operator action, how long will the reactor power level take to reach the point of adding heat (PDAH= 3.3 MW) CO.753 c. When the POAH is reached, how much will moderator temperature change before the temperature coefficient (ALPHA MODERATOR = -1x10E-4 delta K/K/F) causes power to stop increasing.
C1.03 ' QUEST 10N 5.10 (1.50) Briefly explain how a control rod withdrawl of one or two notches can result in a decrease in bundle power? (1.53 (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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3 ' . IEEBd0pyNA5]CS ' QUESTION 5.11 (1.50) State chether the reactor is SUPERCRITICAL, CRITICAL or SUBCRITICAL for each of the followings (assume no operator action and consider only the initial reaction by the reactor) a. Reactor was stable at 10L power, volds fraction just decreased 1% [0.53 b. Reactor is at 25% power and pressure starts decreasing as a bypass valve sticks open.
[0.53 c. Reactor has an infinite period at the point of adding heat (0.53 ' QUESTION 5.12 (1.50) ao During a reactor startup, Keff as.95 when the SRM channels read 100 cps. What will the new Keff be when the SRM channel reads 270 cps? [0.753 b. During a reactor startup, as control rods are pulled to approach criticality, would the time required to reach an infinite period after each rod pull be GREATER or LESS THAN the time for previous rods? Assume all rod pulls are equal.
JUSTIFY your answer.
[0.753 l !
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6 . . QUESTION 4.01 (2.00) a. State TWO (2) conditions that will cause an auto isolation of RCIC. (Include setpoints) (24 0.5) b. List FOUR RCIC turbine trips and include setpoints if applicable.
(4W 0.25ea) . QUESTION 6.02 (3.00) a.
What TWO (2) parameters will initiate the loop select logic f or the LPCI mode of RHR? Include setpoints.
(1.0) b.
Given the following parameters, STATE which loop would be selected by the LPCI Loop Select Logic (A or B) and WHY: (Consider each seperately) 1. Both recirc pumps running at equal speeds, pressure in loop A riser is 900 psig, pressure in loop B is 800 psig.
II.03 2. Both recirc pumps secursd, loop A and B riser pressures are 950 psig [1.03 0UESTION 6.03 (3.00) a.
List FOUR (4) plant conditions with setpoints that are required for ADS to auto initiate.
(1.0) b.
List THREE (3) conditions which will reset the ADS 120 sec. timer (1.0) c.
Briefly explain why it is undesireable to have the Local Control switch for an ADS relief valve at the Alternate Shutdown Panel (156 and 157) in a position other than REMOTE.
(1.0) iOUESTION 6.04 (2.00) With a select Error cn the RWM, a selected rod can still be moved.
Assuse the RWM is not bypassed and no rod block or select errors existed prior to rod selection. Explain the rod movement restrictions, for both and insert and withdrawal, imposed by the RWM.
(2.0)
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3 - l . QUESTION 6.05 (3.00) a. Describe SBGT system response to a loss of a heating element (1.5) duririg operation of the system.
a b. State the reasons for changing the deluge fire systen from automatic operation to manual.
- (1.5) QUESTION 6.06 (2.00) a. Describe the normal and alternate power supplies to the RPS system.
(1.03 b. 1: hat TWO interlocks are associated with the RPS bus power supplies to ensure the power supplies remain redundant and independant from each other.
II.0) . QUESTION 6.07 (2.00) Answer the following TRUE/ FALSE statements concerning the SLC system.
If the statement is false, change the statement so that it is truet a. Proper SLC initiation is verified by at least the followings a ' Loss of Continuity to Squib Valve' annunciator, Squib valve ready light out, pump selected has a red indicator light ON, and Reactor Water Cleanup system isolates.
(0.5) b. SLC pumps and squib valves get their power from the 480V MCC B-17 and B-19 busses (0.5) c. When in operation, the SLC system flow enters the vessel thru the 'B' recirculation loop.
(0.5) d. When the SLC control switch is taken to SYS A position, pump A will start and both squibb valves will fire to ensure a flow path to the reactor vessel.
(0.5) (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
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- QUESTION 6.09 (3.00)
a. You have completed verifing an isolation of the reactor building when a report comes in that instrument air to the-isolation dampers will be lost. What position will the isolation dampers end up and why? (1.0) b. List FOUR signals that will cause a reactor building isolation and an initiation of Standby Gas. (Give setpoints where applicable)(2.0) DUESTION 6.09 (2.00) The reactor is at 70% power when the A Flow Converter fails such that its output is DOWNSCALE. State ALL trips that will occur and all causes for each trip. A center rod was selected at the time of the failure.
[2.03 ' QUESTION 6.10 (1.00) Which of the following will cause the Energency Diesel Benerator to auto-start and close its output breaker: [1.03 a. Dryuell pressure exceeding 2.5 psig b. RPV level lower than -49 inches c. The aux transformer breaker opens with the turbine on-line d Startup transformer output is (90% for 10 sec. and the aux ' transformer breaker is open.
! QUESTION 6.11 (2.00) The plant is operating at 85% power when an operator inadvertently injects HPCI during a surveillance test. Assume NO Scram occurs, that the Feedwater Level Control systen is in THREE ELEMENT CONTROL, and no operator action is taken after the HPCI injection.
Describe the response of the Feedwater Level Control systen to the HPCI injection.
Discuss any effects on reactor water level, feedwater flow, and feedwater regulating valve position.
JUST!FYING EACH. Continue your discussion to a stable condition with HPCI injecting at rated flow.
(2.0) ! i (***** END OF CATEGORY 06 *****) , . - -. .
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B921960sice(_QQNISQ( ,
- QUESTION 7.01 (3.00)
Using the attached curves from E0P-03, answer the followings a. What is the miniaua torus water level for RPV pressure of 900 psig and torus temperature of 165 degrees F? [1.03 ~ b. What is the maximus RPV pressure with a torus water level of 7 feet-7 inches at a temperature of 154 degrees F? [1.03 c. If the torus level COULD NOT be maintained above the Heat Capacity Level Limit, WHAT two insediate actions aust be taken and WHY? [1.0) QUEST 10N 7.02 (2.00) OTHER THAN ALARMS, what FOUR symptons are indicative of a small leak [drywell press. ( 2 psig) inside the primary containment? [2.01 (***** CATEGORY 07 CONTIN')ED ON NEXT PAGE *****)
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PR0g[pVR[$..NQRMA(3.A)NQRMA(3,[M[Rg[Ngy AND PAGE il ) , 8ADig(0919A( 90NJ806 , ) ' , i QUESTION 7.03 (2.00) Using the attached table, state which emergency action level is appropriate for the following conditions: (Choose one EAL for each condition below) 1. unusual event - 2. alert 3. site area energency 4. general energtncy DESCRIPTION EAL (Choose One) .............. ..................... a. Drytell sump flow indicators have exceeded 85 gpa in the last 90 minutes (Reactor at power) (a.)
(0.53 ... .... b. During a heatup both vessel head spray isolation valves malfunction requiring a shutdown per T.S. 3.7.D.3 (b.)
(0.53 ... .... c. RPV pressure spikes to 1100 psig, the plant is manually scrassed, and reactor ... c.)
[0.53 ( po er is 20% with all MSIV's open .... d. PNPS loses all off-site power immediately after an Energency Diesel fails a routine surveillience (Reactor at power) (d.)
(0.51 ... .... QUESTION 7.04 (3.00) If operating in shutdonn cooling and all recirculation flow is lost, PNPS 2.4.24 (Reactor Vessel Cold Water Stratification) states to conitor instrumentation for verification of cold water stratification.
a. What are the indications of stratification? [0.53 b. Other than restorino.nutdown cooling or recirculation flow stratification can sa minimized by af f ecting the flow in two systems. WHAT are these systems are HOW ARE THEIR FLOWS ALTERED to minialze stratification (be specific)? (1.5) c. One concern when recirc flow is lost is the possibility of losing natural circulation. WHAT can be done to ensure natural circulation vill continue? C1.03 (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) - -- _ _
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. B89194QilG86.QQN1BQ( QUESTION 7.05 (2.50) a. List THREE (3) entry conditions for E0r-1 "RPV CONTROL".
Include setpoints.
(1.5) b. A caution in E0P-1 warns the operator that the RCIC turbine gay trip from overpressure in the exhaust line to the torus.
(As low as 37 psig). Assuming HPCI is also in operation and exhausts to the torus WHY is the high pressure not a probles for the oper;iton of HPCl? C1.03 ' QUESTION 7.06 (2.25) Concerning PNPS E0P-2 Failure to Scram Under the sub-heading "Reactor Power", instructions are given to place both ADS inhibit switches in the INHIBIT position prior to injecting boron. Explain why.
(2.25) QUESTION 7.07 (2.00) a. Jet pump failure per PNPS 2.4.23 "Jet Pump Flow Failure" can be indicated when there is a sudden change in _____(1).________, [0.51 ui thout a corresponding change in ______ (2) ________.
CO.5) b. According to Tech Spec 3.6.E and 3.6.F, if a jet pump is considered inoperable (failed) the reactor aust be brought to cold shutdown within 24 hours. Why is jet pump failure such a serious problem.
(1.03 > l l i l (se*** CATEGORY 07 CONTINUED ON NEXT PAGE *****) , _______-_.
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- )UESTION 7.0B (2.00)
a. State the allowable radiation exposure limit in accordance with 10 CFR 20 and PHPS 6.2-001 Radiation Exposure Control Program, assuming the applicant has an NRC Forn 4 and ir 20 years old
, (1.03 cith no previous exposure.
10 CFR 20 PNPS ADMIN LIMIT WHOLE BODY (1),,,,,, ,,,,,,(2),,,,,,,, ,,,, , b. How would this limit change if the applicant did not have an NRC Fora 47 (Answer using PNPS Admin Limits) (0.53 c. Prior to being authorized to exceed 1000 area in a quarter, you first aust determine your lifettee exposure and compare it to your lifetime allowable limit. How is this limit determined? C0.53 , QUESTION 7.,09 (1.50) ' While performing the actions in E0P-1, entered when RPV level ,jggtvessel recched 9 tr,ches, the reactor operator reports,try condition for pressure has exceeded 1095 psig, another he'e'n E0P-1. Relative to E0P performance, WHAT is your next action? C1.53 QUESTION 7.10 (2.25) a. A fire in the main control room (MCR) has caused its evacuation. Prior to leaving, there is time to perform
the immediate actions of PNPS 2.4.143 "Shutdown from Outside the Main Control Roca". Before evacuating the control roo3 and in addition to scramming the reactor, STATE 5 actions I1.253 you would perfore or direct to perfora.
b. If evacuation.ad to be performed prior to scraaning the reactor, hos are you directed to scran the reactor from outside the control (1.03 roos (be specific).
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._88919k99190k.99N]3QL QUESTION 7.11 (2.50) Fill in the following blanks with the maxinua permissible values a. The average rate of change for a normal heatup or cooldown shall not exceed..(1) F when averaged over one hour.
IO.51 .. b. When attempting to open an MS!V, differential pressure across the valve shall not exceed..(2) psid.
[0.53 .. c.-Do not start an idle recircultion pump if the differential teoperature between the idle loop and the operating loop exceeds..(3).. F OR if the differential temperature betceen the vessel steam done and the bottoa head drain is greater than..(4) F.
(1.03 .. da The main condenser vaccoun pump aust be secured proir to the reactor exceeding..(5).. % thermal power.
[0.5) ,
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3 . QUESTION 8.01 (3.00) Tho reactor is operating at a high power level when a TURBINE TRIP occurs causing the reactor to scram on HIGH FLUX. Answer the followings a. Has a violation of SAFETY LIMITS occured? If so whyl if not, why not? [1.5) b. What insediate actions are you as Watch Engineer required to take? (2 required) (1.5) QUESTION 9.02 (1.50) According to Tech Specs section 6.6, which of the following events requires prompt notification (within one hour) of the NRC (more than one answer any be required): [1.53 a. Malfunction of a Reactor Building air lock such that the ability to have at least one door shut is not possible (at power) b. Initiation of a CSCS system caused by operator error during survellience testing c. A calculated reactivity balance indicating a shutdown margin less conservative than specified in Tech Specs, d. Plant shutdown required by a limiting condition for operation.
QUESTION 0.03 (1.00) State the PNPS SAFETY LIMIT (s) for cold shutdown conditions.
II.03 (***** CATEGORY OB CONTINUED ON NEXT PAGE *****)
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.. GUESTION 8.04 (2.00) The reactor is at 100% power and 100% recirculation flow with the tee recire pumps operating at speeds within 5% of each other.
During a routine surveillance, it was found that indicated core flou was 12% greater than established total core flow derived from recirc loop flow. The ' diffuser to lower plenus differential pressure' has not_ varied from the established jet pump operating characteristics.
Using the attached Technical Specifications, can the plant continue to operate under these conditions? -If so why, if not why not? Reference any sections of Tech Specs used in your answer.
(2.03 ,
QUESTION ~8.05 (3.00) a. According to PNPS 1.4.5 ^PNPS Tagging Procedures" who (by title) is authorized to approve tagout sheets so that protective tags can be put in place? C0.53 b. For what two types of plant systems is an independent second
verification of tag placement required? [0.53 ' c. TRUE or FALSE: (if FALSE, correct so that it is TRUE) l. naster Danger tags are used to isolate major components (e.g.
, pumps, compressors,etc.) where Red Tags are used for minor components (e.g. switches, valves, fuses, etc.)
[0.53 2. Caution tags are suited for components like the SLC control switch in the main control room when one of SLC pumps is out for maintenence and the other is ' required for operation (0.53 d. What must be done to ensure tags temporarily removed from a
caintenance boundary (isolation) are documented and who(by title) is i responsible for for making this documentation? [1.01
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' . SUESTION 8.06 (3.00) While performing core alterations during a refuel outage, a surveillance is performed on the Standby Gas Treatment System.
You note in your shift log that painting is in progress on several s levels of,the Reactor Building. It is reported to you that SGTS ' train A fan failed the flow test. Using the attached Tech Specs, ansrer the following: - (be sure to include any reference to Tech Specs as part , of your answer) Can present and future refueling operations continue? 'If not, why notl if so, under what restrictions? (3.01 SUESTION B.07 (2.50) 1.
The PNPS MINIMUM crew composition per Tech Spec Table 6.2-1 shall be as follows: (1.5) ( ... a)... OPERATING: SR0(s) ... b)... ( R0(s) UNLICENCED OPERATOR (s) ...(c)... . ................................................................. REFUELING: SR0(s) ...d) ( ... R0(s) (e)... ... UNLICENCED OPERATOR (s) ( f )... ... II. Co:plete the following concerning the FIRE BRIGADE: (1.01 A fire brigade of...(a) neabers are required to be onsite .... (0.5) per Tech Spec at all times. This excludes...(b).... seabers (0.5) of the operating staff, necessary for a safe shutdown t l' (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) - - _--.-_.
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32._6ED]h]!!B61]yg_BBggggUBEg3_gghg]Ijghg3_gyg_(jgjl61]QUS PAGE 18 . $UESTIDH B.08 (2.50) l Ccncerning control rods: .a. Tech Specs states that a maximum of eight control rods can be inoperable while at power. What is the basis, per T.S. BASES 3.3 ano 3.4, for this number? [1. 5 3 ' b. TRUE or FALSE; ^ ' A fully inserted control rod that has been electrically i disarmed is considered inoperable.
' [1.03 IUESTION 8.09 (2.00) ' According to procedure PNPS 1.3.6 "Adherence to Technical Specifications"
- elth proper consideration and authorization, licensed operators can knouingly deviate from License Conditions or Tech Specs.
a. To deviate from tech specs without a "Tech Spec Clarification", personnel censidering protective action aust consider two criteria first. What are the two criteria.
[1.03 b. Who(by title) may approve this action, as a minlaus, and what additional action must be taken, if time permits, prior to initiating the protective action.
[1.03 fuESTION 0.10 (2,00) A Maintenance Request is brought to you for approval. What actions or checks will you perf orm as Operating Supervisor to approve the request.
(State 5 and assume the request is valid) l2.03 l l
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t t . QUESTION 8.11 (2.50) a.
It is a precaution in PNPS 4.3 Fuel Handling, to ensure that .certain types of fire fighting equipment are not used to fight fires in the vicinity of the new fuel storage area. WHAT types (2) .cof equipment is prohibited and WHY? (1.5) . b. As SRO for refuel operations, you are observing the renoval of a fuel bundle from the core when it is reported to you that vessel water level is decreasing due to an unisolable leak. In accordance with PNPS 4.3 Fuel Handling, you should: [1.0) (choose one) 1.
insert it back in the core 2. remove it and transfer it to the spent fuel pool 3. OR suspend all fuel novement until the vessel level is restored (***** END OF CATEGORY 00 *****) (************* END OF EXAMINATION ***************) .- _ _ _ ,... . _,.. . - _ - - _ -.. __,,,- -
. 5.~ THEORY OF' NUCLEAR POWER PLANT QPERATigN _FLUlDS,AND PAGE 20
i . ISE8dODYNAd[QS ~ ANSWERS -- P!L6 RIM-87/12/07-GRATTON, C.
ANSWER 5.01 (3.00) a.
increase (0.253, as the pressure of the system decreases the flow increases due to the centrifugal pump head / flow characteristic.
- (0.753 b. decrease (0.253, as the pressure of the system decreases the operating point on the pump characteristic curve is shifted to a lower pump discharge pressure.
[0.75) c. re:ains the same (0.253, pump speed is independant of discharge pressure (or dependant on the electrical characteristics of the pump 3 (0.75) REFERENCE G.E. Heat Transfer and Fluid Flow, Ch. 6 pg. 6-95 & 6-96.
K/A's... 293006 K1.08 2.5/2.6 291004 K1.05 2.8/2.9 291004K105 293006K108 ...(KA*S) ANSWER 5.02 (2.50) a. LH6R [0.53 b.
APLHGR or MAPLHGR (0.51 l c. FM-f uel clad cracking due to dif f erential expansion of the pellet and the cladding (0.5) LC-pin power is limited to prevent greater than or equal to 1% plastic strain on the cladding.
(0.5) d. Transition boiling occurs which can result in clad cracking ! (due to loss of adequate cooling) (0.53 REFERENCE ' G.E. Heat Transfer and Fluid Flow, ch.9, pg. 9-15, 9-18, 9-19 K/A's... 293009 K1.10 3.3/3.7 293009 K1.12 2.9/3.5 293009 K1.07 2.8/3.6 293009 K1.08 3.0/3.4 293009 K1.21 3.1/3.6 ! l 293009K107 293009K108 293009K110 293009K112 293009K121 l...(KA'S) L i
. 5 _,18E081.0E_NQC(E@@_&QMEQ_E(@N1_Q&E8@llqN _E(ylqS,@NQ PAGE 21 i
i INE85QQ1N@QlCS , ANSWERS -- PILGRIM-87/12/07-6RATTON, C.
' ANSWER 5.03 (2.50) a. Rod A (0.5) Upon scram recovery, fission product poisons cause a severe flux depression in what was the highest power producing region of the care. This results in a higher relative flux in regions of low poison concentration. These shifts in the flux distribution increase the worth of peripherial rods and decrease the worth of those in the center of the core. [1.0) b. Shallow- [0.5) Are (0.53 REFERENCE G.E. Reactor Theory, ch.5 pg. 8,9,18,25; Ch. 6 pg. 12 K/A's... 292006 Kl.07 3.2/3.2 292006 K1.09 2.5/2.5 292005 K1.10 2.8/3.3 292005 K1.12 2.6/2.9 292005K110 292005K112 292006K107 292006K109 ...(KA'S) ANSWER 5.04 (2.00) a) False (0.25) Equilibium level at 50% is approximately 2/3 the equilibrium level at 100% power (0.25) b) True (0.5) c) True (0.5) d) False (0.25) Xenon has a half-life of about 9.2 hours (0.25) REFERENCE G. E. Reactor Theory Manual, pages 6-8, 6-15, 6-11 and 6-13 Enabling objective 2.1, 2.5, 2.6, 3.5, 3.6, Chapter 6 K/A's... 292006 K1.03 2.9/2.9 292006 K1.13 2.6/2.6 292006 K1.04 2.9/2.9 292006 K1.07 3.2/3.2 292006K103 292006K104 292006K107 292006K113 ...(KA'S) , i l ANSWER 5.05 (1.50) a) To prevent cavitation in the condensate pumps (or to provide adequate condensate pump suction head) (1.0) b) A plant operating with excessive condensate depression reduces plant efficiency thus shortening the fuel cycle (0.5) REFERENCE G. E. Heat Transfer and Fluid Manual, page 7-45 K/A's... 293004 K1.12 2.9/3.1 293007 K1.09 2.5/2.7
. 5 _ lh[QBY,QE_NQC({QB EQWEQ E(@NI_QEE88[[QN _E(Q1QQ3,6NQ PAGE 22 t i . ISEgqqQ1N@QlC@ ANSWERS -- PILGRIM-87/12/07-GRATTON, C.
293004K112-293007K109 ...(KA'S) ANSWER.
5.06 (3.00) a) Reactivity increases (0.25) the cooler water entering reactor causes increased neutron moderation (and contact time with the f uel shif ting the boiling boundary up in the core).
(0.75) b) reactivity increases (0.25) due to collapse of voids, increase in moderator density, with more thermalization (0.75) c) reactivity decreases (0.25) due to higher fuel temperatures acre parasitic absorptions (doppler effect) occur in the resonance region (0.75) REFERENCE Ge E. Heat Transfer and Fluid Flow Manual, pages 9-16, 9-19 and 9-33.
Enabling objective 3.1, 3.6, 4.1, 4.2, 5.1 and 5.3, Chapter 9 KiA's.a.
292004 K1.01 3.2/3.2 292004K101 ...(KA*S) ANSWER 5.07 (2.00) a. 4 (1.0) b. 3 (1.0) REFERENCE G.E. Heat Transfer and Fluid Flow Manual, Chapter 6, pages 73 thru 77, 97,108 and 109 need ref document K/A's... 291004 K1.06 3.3/3.3 291004 K1.07 2.8/2.8 291004K106 291004K107 ...(KA'S) ANSWER 5.08 (1.50) I b. 70 F/hr (1.5) REFERENCE ScE. Heat Transfer and Fluid Flow Manual, Objective 2, Chapter 3 L
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IBEQ5QQ1NQQlC@ . ANSWERS -- PILGRIM-87/12/07-GRATTON, C.
K/A's... 293003 K1.23 2.8/3.1 1293003K123 ...(KA*S) ANSWER 5.09 (2.50) a. Period = Doubling Time x 1.443 = 90 sec. x 1.443 = 130 seconds (0.75) b. P(f)= P(i)eE(t/T) 3.3x10E6 / 2000 = eE(t/T) g?lh(7.4)= t/130 q s W to 963 sec. or 16.03 min.- (0.75) c. reactivity =.0072/((.1 x 130) + 1) =.0005 delta K/K delta T = alpha T x reactivity added-1x10E-4 x.0005 deltaK/K/F = 5 F increase in temperature ( 1. 0 ) = (note to grader-part c is to be graded on methodology) REFERENCE 6.E. Reactor Theory Manual, Chapter 5 Objective 2.7, Chapter 4 Objective 2.5, Chapter 3 Objective 3.5,3.6,3.7 K/A's... 292003 K1.05 3.7/3.7 292003 K1.08 2.7/2.8 292003 K1.09 2.5/2.6 292003K105 292003K108 292003K109 ...(KA'S) ANSWER 5.10 (1.50) The steam bubbles generated by the withdrawl of a shallow rod increase the void fraction [0.75), which adds negative reactivity off-setting the positive reactivity effects of the rod withdrawl (0.75) REFERENCE GE Reactor Theory, Chap. 5, pp.16,17,18 K/A's... 292008 Kl.19 3.1/3.2
! i , _ . _. ., _, ,, ,. . .. _ _ _. ,,. . . -
- . 5i__IBEQB1_Q[_NQC(E68_EQMEB_E(6NI_Q&EB@l[QN _E(QlDSs_AND PAGE 24 i ..IBEBdQDYN85[CS . ANSWERS -- PIL6 RIM-87/12/07-GRATTON, C.
ANSWER 5.11 (1.50) a. supercritical (0.53 b.
subcritical (0.53 - c. critical (0.53 REFERENCE GE Reactor Theory Chap. A pp. 33,34 K/A's... 292000 K1.15 3.7/3.7 292000 K1.22 3.5/3.6 292008 K1.20 3.1/3.2 ' ANSWER 5.12 (1.50) a.
CR1(1 - Keff1) = CR2(1 - Keff2) (0.25) CR1/CR2(1 - Keffi) = (1 - Keff2) (1 - Keff2) (0.25) -100/270(1 .95) = .0185 = (1 - Keff2) Keff2 = 0.9815 (+/-.002) (0.25) b.
Greater (0.251.
The fractional change in neutron population per generation becomes less as you approach criticality (Keff=1).
Therefore, it requires a longer period of time for neutron population to stabilize (0.5).
Also accept answers which explain that there are more generations t's go through prior to population stabilizations.
REFERENCE REACTOR THEORY REFERENCE TEXT CHAPT!R 3 pg 3-5 to 3-15.
REACTOR THEORY STUDENT GUIDE ELO 1.3, 1.5 292003K101 292008K104 ...(KA'S)
- . 6 __P($$1,}Y@IED),9E!]GN CgNIB9L,6ND,jN@lBUDENI@IjgN PAGE. 39 t
3 ANSWERS -- P!LGRIM- -87/12/07-CRATTON, C.
ANSWER '6.01 (2.00) 4.
-Reactor pressure' low, 50-100 psig.
-Hi steaa' supply delta p, 300% for >3 sec-Hi turbine area or steam line temperature, 150-200 F in torus, RCIC roca, RCIC valve roca (any 2, 0.25 for trip,0.25 for setpoint) b. -Turbine overspeed 9 125% -Loo suction pressure 754 vaccuum V3 " -Hi turbine exhaust ~ pressure; 25 psig-Auto isolation signal-Manual pushbutton at the 904 panel-Manual trip at the turbine (any 40 0.25ea) REFERENCE PNPS STUDENT' GUIDE 09-04, Enabling Objective 11,13 .K/A's... 217000 K4.04 3.0/3.1 217000K404 ...(KA'S) ' ANSWER 6.02 (3.00) a.
Drywell press >= 2.5 psig C0.5) and RPV water level <= -49"(0.5) b.
1. Loop Ato.5), Lower pressure in loop B indicates break is most probably located in that loop, logic selects opposite loopCO.5) 2. Loop B(0.5), logic defaults to Loop B when no clear indication of-a break is apparent (0.5).
REFERENCE PNPS STUDENT GUIDE RHR-LPCI Learning Objective 8,9,10 K/A's... 203000 Kl.13 3.9/4.0 203000 K4.11 4.0/4.0 303000K113 203000K411 ...(KA'S) ' . _ _ . _. _., _ _. _ _ _ _ _ _ -. _ _ _ _ _,,, , --
_ _ - _ . .__ - . . 6..PL6hl_SYSIED}_9E!!GU3_CgyIB9L 6hD_lhSIBudEUISIlgd PAGE 26
3 AkSWERS--- PILGRIM-87/12/07-GRATTON, C.
ANCWER 6.03 (3.00) a.
Drywell pressure >= +2.5 psig.
(0.25) Reactor water level (= -40 inches.
(0.25) 120 second timer timed out.
(0.25) Any Low Pressure CSCS pump running with >= 150 psig discharge pressure (0.25) b.
A loss of power Ticer reset button being depressed.
Rx water level increasing to > -49 inches within 120 seconds Manually resetting the 2.5 psig Drywell Trip after the initiating signal has cleared (3 0 0.33 ea' c.
In a position other than REMOTE, the valve will not respond to an ADS automatic initiation signal (0.5) or to a manual signal from 'the control roon (panel 903).CO.53 REFERENCE PNPS SYS MANUAL ADS-Enabling Objectives 5,13,14,15,18 K/A's... 218000 K1.05 3.9/3.9 210000 K5.01 3.8/3.8 218000 K4.01 3.7/3.9 218000 K4.03 3.8/4.0 213000K105 213000K501 218000K401 218000K403 ...(KA*S) ANSWER 6.04 (2.00) It can be moved out one notch before a withdraw error will block further movement (1.03. If the rod is inserted, it will move as far as the operator wants (to full insertion as long as it is not the third insert error) [1.0).
REFERENCE PNPS Student Guide 06-03, Enabling objective 25,28 LG-5 K/A's... 201002 A2.03 2.9/2.8 201002 A2.04 3.2/3.1 201002 K4.02 3.5/3.5 201002A203' 201002A204 201002K402 ...(KA*S)
_ _ _ ._ _ . . .6..P($$1,SYSlEUS_ DES!GMs_ CON 180(s,6ND_!US18ydEH16110N PAGE 27
A'NSWERS -- PILGRIM-87/12/07-GRATTON, C.
ANS'2ER 6.05 (3.00) a.'If a heater trips or burns out a current sensing relay will . trip the SBGT system blower for that train.(0.5) If it were allowed to operate without heaters, the moisture content would increase and decrease th efficiency of the charcoal' filters. (1.0) .b.
Manual operation of the deluge system eliminates the possibility of an automatic initiation of the deluge system (af ter a design basedaccident(whichcouldwashradio-iodinesintotheTBfloor drains) (1.5) REFERENCE PNPS Student Guide 08-03 Learning Objective 118,14D,11D PNPS Misc Natis PDC-86-70 Learning Objectives 1, 5 K/A's... 261000 K4.02 2.6/2.0 261000 K4.03 2.5/2.7 261000 K4.06 2.4/2.6 261000 K5.01 2.3/2.6 261000K402 261000K403 261000K406 261000K501 ...(KA*S) ANSWER 6.06 (2.00) a. RPS MG sets A and B [ Busses B-23 and B-22 is also acceptable) (0.5) provide normal power and both busses receive alternate power from .MCC-10 (0.5) b. - An RPS buss cannot be powered from both its normal and alternate power supplies at the same time (0.5) - RPS A and B cannot be powered from their alternate power supplies simultaneuosly (0,5) REFERENCE PNPS Student Guide 07-02 RPS, Enabling Objective 13,14 K/A's... 212000 K2.01 3.2/3.3 212000 K4.01 3.4/3.6 212000 K4.04 3.1/3.1 212000K201 212000K401 212000K404 ...(KA*S) l l l l .,..... -- ..- - _ - - -
. . 6,,&(@81 SYSIEUS DES!GN _ CON 18Q(t,8ND,18S18UQE816110N PAGE 28 t t ASSWERS -- PIL6 RIM-87/12/07-GRATTON, C.
ANSWER 6.07 (2.00) a. T (0.5) b. T (0,5) c. F (0.25)SLC has its own piping to the core plate region (0.25) d. F (0.25)0nly 'A' squibb valve will fire (0.25) REFERENCE .PNPS Student Guide 06-06 SLC, Enabling Objectives 4,7,8,11,12,14 K/A's... 211000 K1.06 3.7/3.7 211000 K4.07 3.8/3.9 211000 K4.08 4.2/4.2 211000K106 211000K407 211000K408 ...(KA'S) -ANSWER 6.08 (3.00) a.-The valves will remain shut (0.25) because they fail shut upon loss of instrument air. (0.75) b. Reactor Low Level (+9") High Drywell Pressure (+2.5 psig) High Refuel Floor Rad Levels (>16 ar/hr) Four Downscale Refuel Floor Exhaust Duct Radiation Monitors Tuo Downscale combined with One Upscale Refuel Floor Exhast Duct Rad Monitors (49 0.5ea) %% of noV% go n e. l d Y. 2, REFERENCE Student Guide 08-05 Plant Ventilation, Enabling Objectives 5,6,7 K/A's... 288000 Kl.02 3.4/3.4 288000 K4.01 3.7/3.9 288000 K4.02 3.7/3.8 288000 K6.03 2.7/2.7 289000K102 288000K401 288000K402 288000K603 ...(KA'S) ' boss * ^'" i AC (Y-I') ' p. E 1. 4 - 1.
haps %dse 514
l ,. - -. , - _-. - ._._..., . ..- .. - ,, -,. -,. .
__ .. . is..g(QN1_SISIEMS, DES!6H _CONIE0(t_@ND,[NSIRyMENI@l10N PAGE 29 i ANSWERS -- PIL6 RIM-87/12/07-GRATTON, C.
' ANSWER 6.09 (2.00) ROD. BLOCKS - flow comparator sisaatch (>10%) or F. C. Lo p Tr ip (0.5) - flow biased trip (power >.58W + 50) (0.5) - RBM flow bias trip-(0.5) HALF SCRAM CHANNEL A - flow bias (power >.58W + 62) (0.5) REFERENCE PNPS Student Guide 07-01 NMS pp LG-12,13,14,15 K/A's... 212000 K6.02 3.7/3.9 215005 K1.10 3.3/3.3 215002 K1.01 2.9/3.0 215005 K1.16 3.3/3.4 215005 K6.07 3.2/3.3 212000K602 215000K110 215002K101 215005K116 215005K607 ....(KA*S) ANSWER 6.10 (1.00) , da.
[1.01 REFERENCE PNPS Student Guide 01-02 AC Distribution Enabling Objective 13 K/A's... 262001 kl.01 3.0/4.3 262001 k3.02 3.0/4.2 262001 k4.03 3.1/3.4 262001 k6.03 3.5/3.7-263001K101 26200lK302 262001K403 262001K603 ...(KA'S) _ ANSWER 6.11 (2.00) RPV level would increase due to the extra HPCI injection flow (0.4).
This flow is not sensed by the FWLCS so the FRV will not immediately reposition (0.4).
As RPV level increases, a level error signal will develop which results in FRV partially closing (0.4).
Level will stabilize at a point high enough where the level error signal compensates for the HPCI injection flow (0.4).
Total feedwater flow will decrease by the amount of HPCI injection flow (0.4).
REFERENCE PNPS Feedwater Level Control Reference Text pg. 2-2 PNPS Feedwater Level Control Student Guide pg. LG-23, LG-24. ELO 38, 40.
- - - -- - . -.- .- -. .. , _. _ - - - - _ -,. -.
h _EL64I_gySIggS_DESigy,,gggyg9bi 6dE idSI69dE8Igl{gg PAGE 30 A$SWERS -- PILSRIM-87/12/07-6RATTON, C.
259001A101 25900lK102 ,,,(gg.g3 . I
. . Zs. 88QQEDQ8ES_ _N0856(t_QBNQ856(g_EdE8EENQY_@ND PAGE 31 . 86Q10(0@lq@(_Q0N180( ANSWERS -- PIL6 RIM-87/12/07-GRATTON, C.
ANSWER 7.01 (3.00) a. 103 inches (8 feet 7 inches] (1.03 tol t 3" b. 800'psig (1,03 2 le peg - c.
If suppression pool level cannot be maintained, manually scras the reactor (0.333, and depressurize the RPV using ADS CEOP-073 (0.333.
This is to ensure that condensation of the steam in the vessel in the event of a LOCA can be assured.to.333 REFERENCE PNPS Technical Bases of E0P's Primary Containment Control-Sect. !!!.D K/A's... 223001G014 3.8/3.7 223001G014 ...(KA*S) ANSWER 7.02 (2.00) -Excessive sump pump operation (due to increased leakage to the drywell equipment drain sump or drywell floor drain susp.)
[0,5) -Abrupt change in drywell humidity (as indicated on panel C85 recorders) (0.53-High radiation detected on drywell leakage detection systea (panel C19.3 (0.53-Significant pressure changes as recorded on drywell pressure recorders.
[0.53 - s p ti u,* 4 4 ~9 c%e s c..w < a.A r AM mu e ,tu.<%,, q.. i e,5 e.], REFERENCE PNPS procedure 2.4.14 Leakage Inside the Primary Containment PNPS Tech Specs 3.6.C.1 ANSWER 7.03 (2.00) A. 2.
[0.53 B. 1 (0.53 C. 3 (0.53 D. 1 (0.53
. it._88QCEDyBES_ _N08d&(t_8@NOBd8(3_EdE8@ENCy_AND PAGE 32 88D10(OS!C8(_ CON 18Q( . ANSWERS -- PILGRIM-07/12/07-6RATTON, C.
REFERENCE PNPS Emergency Categories and Associated EAL's 5.7.1.1 App. E . ANSWER 7.04 (3.00) . a. A large differential temperature between the Reactor vessel bottom and elsewhere in the vessel (as read on recorders TR263-105 and TR263-104).
I0.53 b. RWCU and CRD (0.53-Throttle down on FWCU suction (M0-1201-853 from the recirculation loop (keep RWCU suct. press >20 psig) -Reduce CRD flow [2 at 0.5 ea) con o r" C
- /9'*8U W di t*9 c. Raising vessel water level (to +46 inches)(instrument zero)
ensures natural circulation will occur II.0J.
REFERENCE PNPS Procedure 2.4.24 Cold Water Stratification PNPS Procedure 2.4.25 Loss of Shutdown Cooling K/A's... 295021 K1.02 3.3/3.4 295021 K1.04 3.6/3.7 295021 K3.01 3.3/3.4 295021 al.01 3.4/3.4 295021A101 295021K102 295021K104 295021K301 ...(KA'S) ANSWER 7.05 (2.50) a. ( +9 inches vessel level >iOB5 psig vessel pressure - > 2.5 psig drywell pressure Conditions for a scras met AND reactor power >3% OR undeterminable Cany 3; 9 0.53 b. HPCI exhaust turbine trip is set considerably higher (150 psig) than RCIC turbine exhaust trip and is therefore not a concern.
[1.01 REFERENCE PHPS E0P-1 RPV Control PNPS Technical bases for the E0P's - Cautions K/A's... 206000 K4.01 3.0/3.9 206000 A2.16 4.0/4.1 295024 Gen 11 4.3/4.5 206000A216 206000K401 295024G011 ...(KA'S) . -- .- ,, . _, ., -, -,. _ _ _ _ - _ _ _ _, _ _,. , .,,, - - - - -
. Zt__P8gqEQQBES _N085@(g_8BN0806(3 EdEBGENQ{ QNQ PAGE 33 889196991C@(_Cg818g( . ANSWERS -- PIL6 RIM-87/12/07-GRATTON, C.
ANSWER '7.06 (2.25) ADS initiation with CSCS systems injecting will result in a large amount of cold, unborated water being injected into a reactor that is critical or shutdown on baron. [0.75) The positivs reactivity addition due to the boron dilution and the temperature reduction [0.753 may result in a power excursion large enough to cause core damage [0.75).
REFERENCE PNPS E0P-2 Failure to Scram K/A's... 295037 K2.10 3.8/4.1 295037K210 ...(KA'S) ANSWER 7.07 (2.00) a.
(1) cere differential pressure, recire flow, OR jet pump flow indication (ONLY ONE REQUIRED) (0.53 (2) recirc pump speed [0.53 b. Loss of a jet pump could preclude the capability to mLintain 2/3 core coverage during a LOCA. Eh0F C.o.53
ed wM inu cce \\ana de,e m (A (o.5 ) PNPS 2.4.24 JET PUMP FLOW FAILURE PNPS TECHNICAL SPECIFICATIONS 3.6.EhF, and Bases K/A's... 202001 A2.01 3.4/3.9 202001 Gen 6 3.0/4.1 202001 K5.02 3.1/3.2 202001A201 2020016006 202001K502 ...(KA*S) ANSWER 7.08 (2.00) a.
(1) 3000 area /qtr (0.53 (2) 1000 mres/qtr (0.5) b. The applicant would be limited to 250 mrea/qtr IO.5) c. Lielt = 5(N-18) Ree where N equals the applicants' age in years (0.5) REFERENCE PNPS 6.2-001 Radiation Exposure Control Progrin K/A's... 294001 Kl.03 3.3/3.8 294001K103 ...(KA'S) -- . . . .. .-... -. .,
. Zt..P80CEDQ8ES_,N085@(1,8BNQBd@(i,EMEBGENC{,ANQ PAGE 34 ,-80219LQGIC@(_CONIBQ( HNSWERS -- P!LBRIM-87/12/07-GRATTON, C.
ANSWER 7.09 (1.50) Additional entry conditions for a procedure initially entered which subsequently occurs, requires that the procedure be re-entered at the begining. (1.53 - REFERENCE PNPS Student Guide - EDP'st Structure, Format & Use of, 01-02 .K/A's... 259002 Gen 11 3.3/4.0 259002G011 ...(KA'S) ANSWER 7.10 (2.25) a. -Announce the event over the PA-Reduce recirc flow to minimum ' -Wait 20 seconds then trip the turbine-Verify all rods are inserted-Verify generator output breaker is open-Trip the feed pumps tany 59 0.25 ea) b. Open the APRM A & B power supply breakers (on RPS panel C511) (1.03 REFERENCE PNPS 2.4.143 Shutdown from Outside the Main Control Roon K/A's... 295016 Gen 10 3.8/3.6 295016 A1.01 3.8/3.9 295016A101 2950166010 ...(KA*S) ANSWER 7.11 (2.50) 1.
100 NOS^ > 9 "#' *3 2. 200 3. 50; 145 4. 5 [ecch correct answer 0.53 REFERENCE PNPS 2.1.9 Reactor Recirculation Pump Operation, Precaution A PNPS 2.1.3 Hot Standby Maneuvers Startup with MSIV's Closed PNPS Tech Spec 3.6.A PNPS 2.2.92 MS Line Isolation and Turbine Bypass Valve PNPS 2.2.93 Main Condenser Vaccuum System K/A's... 239001 K4.09 3.3/3.3 239001 A4.01 4.2/4.0 239001 Gen 5 3.2/4.2 239001A401 2390016005 239001K409 ...(KA'S) . _ _ _ __ _ _.
.
. . et__6951N! SIB 8IlyE_P89CEgyBES _CQNp!Ilghg3_6Np_LidlI61]QNg PAGE 35
AASWERS -- PILGRIM-87/12/07-GRATTON, C.
' ANSWER 8.01 (3.00) a. Yes, a violation has occurred. [0.53 A, Safety Limit is assumed to have been exceeded if a scram is initiated by means other than the primary source signal (in this case, the control valve fast closure).tl.0) . b. Shutdown the reactor Notify the NRC within one hour (29 0.75 eaJ ' REFERENCE .PNPS Technical Specification 6.7 and 1.1.C SRO Augmentation 01-04 Objective 0 K/A's... 295006 Gen 11 4.5 295006 Gen 4 4.2 295006 Gen 2 4.5 295006 Gen 1 4.1 295006G001 295006G002 295006G004 295006G011 ...(KA'S) ANSWER 8.02 (1.50) , a,c,d (3 0 0.5 ea) REFERENCE PNPS Technical Specifications 6.6 K/A's... 292002 Gen 3 3.7 294001 A1.15 3.4 292002G003 294001A115 ...(KA'S) ANSWER 0.03 (1.00) RPV level in cold shutdown shall be maintained > 12 inches above normal TAF.
II.01 REFERENCE PNPS Technical Specifications 1.1 and 1.2 SRO Augmentation 01-02 Objective 1 K/A's... 295002 Gen 6 3.1/3.9 292002G006 ...(KA*S) . - _ -.- .... . .. _
. . 't.. 0051N! SIB @l[yE &BOCEQURES _CQNQ[l[QNSg_@gQ.([dll@l[QUQ PAGE 36 t AhSWERS -- PIL6RIN-87/12/07-GRATTON, C.
ANSWER 8.04 (2.00) Yes; [1.03 per Tech Spec 4.6.E. operation may continue as long as two or more of the listed indications (T.S. 4.6.E.1-3) of jet pump inoperability do not occur simultaneously. [1.03 . REFERENCE PNPS Technical Specifications 3.6.E and 4.6.E SRO Augmentation 01-03, Objective 3 K/A's... 202001 A1.03 3.6/3.6 202001 A1.01 3.6/3.5 202001 Gen 6 3.6/4.1 202001A101 202001A103 202001G006 ...(KA*S) ANSWER 8.05 (3.00) a. Watch engineer or Operating Supervisor (on-duty) has authorization for issuance (0.5]; b. An independent verification is required on Nuclear Safety Systems or Auxiliary systems that support Nuclear Safety Systems [2 0 0.5ea3 c.
1. FALSE CO.1); Master tags are used to keep control room and canagement personnel cognizant that maintenance is being performed on a system (0.23. Red Tags are attached to a switch cr valve to prohibit the operation of the device, to protect personnel from injury and equipment from damage (0.23.
2. TRUE to.53 d. The maintenance supervisor [0.53 aust provide a written reason for the tags removal in the comment block on the tagout sheet for the tags to be removed [0.5).
REFERENCE PHPS 1.4.5 PNPS Tagging Procedures X/A's... 294001 K1.02 3.9/4.5 294001K102 ...(KA*S) ict;; Mt u, .,,s ,,c, n.e. u m <
^! __8H51 BIS 18@llyE_&BQCEQy8ES,CQUQ!!!Q8Qi_@8Q_([dll@l[QNQ PAGE 37 i A'NSWERS -- PILGRIM-S7/12/07-GRATTON, C.
ANSWER 0.06 (3.00) The transfer in progress may continue.[0.5) Any additional alterations are limited by T.S.3.7.B.1.e and T.S.3.7.B.1.c which state: "refuel operations cay continue for the next 7 days provided the other train is demonstrated to be operable within 2 hours and daily thereafter.[1.5) In addition, tests and analysis in T.S. 3.7.B.1.b.2 shall shall be performed on the operable train following its operational test, as required by T.S. 4.7.B.1.a.3 [1.0) REFERENCE PNPS Technical Specifications 3.7.B.1, 4.7.B.1 SRO Augmentation 01-04, Enabling Objective 3 K/A's... 261001 Gen 6 2.6/3.7 234000 Gen 6 2.5/3.7 234000G006 261001G006 ...(KA'S) ANSWER 0.07 (2.50) !. a. 2 d.
I b. 2 e.
c. 2 f.
(6 9.25 ea) ....._____.....___....__...______..__________........._____ .!!. a. 5 b.
(2 0.5 ea) REFERENCE PNPS Technical Specifications Table 6.2-1, Sect. 6.2.6 PNPS Procedure 1.3.34 Conduct of Operations SRO Augmentation 01-04 Enabling Objective 5 K/A's... 294001 K1.16 3.0 294001K116 ...(KA'S) . -.. - -.. - -. .. - -.. - -.. _, - .. -. - - . _ - - . .
.. _ _. - _ _ _ . . ac._8 QUIN [ SIB @l[yE_PRQCEDyRES,CQND[l[0N{i_6NQ (idl1@llgN@ PACE 38 t AhSWERS -- PILGRIM-87/12/07-GRATTON, C.
ANSWER 8.08 (2.50) a. CParticularly late in life, shutdown margin can be assured with more than 8 inoperable (partially withdrawn) rods); a core experiencing cere than 8 inoperable rods is-indicative of a generic rod drive problem and the reactor is' shutdown. C1.53 ""
(WM Att<pr othsuon oE conte woo *.ig foax as * ** of *f " E b.
false II.0) REFERENCE PNPS Technical Specifications 3.3.A and C, Bases 3.3 and 3.4, Definition E SR0_ Augmentation 01-01, Enabling Objective 3.d 01-03, Enabling Objective 4 K/A's... 201003 Gen 63.3/3.9 201003 Gen 5 2.8/3.7 . ANSWER 8.09 (2.00) a. -It is necessary to protect the health and safety of the public-No action consistent with the lic'snse or Tech Specs that can provide adequate or equivalent protection is immediately apparent [2 0 0.5 es3 b. As a minimum, a licensed SRO aust approve the action [0.53 Notification by phone to the NRC PRIOR to the action taking place IF TIME PERMITS. (0.53 REFERENCE PNPS 1.3.6 Adherence to Tech Specs PNPS Tech Specs Sect 6 294001K116 ...(XA'S) i , i l ! ! !
_ _ _ _ . . .
- li._8td[NIQ18@l[yE_&BQCEQQ8ES _CQNQ{110NS _@NQ_([dlI@l[QNQ PAGE 39-t t
dNSWERS -- PILGRIM 87/12/07-GRATTON, C.
- ANSWER 8.10 (2.00) Any 5 are acceptable (5 9 0.4 ea) -Check for duplication-Check for Tech. Spec involvement-Check MR for completeness-Assign MR number from the WE's "Maintenance Request Log" -Assign a priority-Identify a need for isolation-Identify a need for a RWP-Sign the MR approving the work REFERENCE PNPS Procedure 1.5.3 Mainenance Requests, VI.2 K/A's... 294001 K1.06 3.4 294001 K1.07 3.6 294001K106 294001K107 ...tKA'S) ANSWER 8.11 (2.50) ao Foaa and sist spray fi e fighting equipmentt0.53 are not to be used in the vicinity a the fuel storage area because of the risk that their use may af f ect the Kef f of the new f uel.
[1.03 b.
1.
(1.03 REFERENCE ' PNPS 4.3 Fuel Handling, Precaution G and Y X/A's... 234000 Gen 10 2.9/3.5 234000 Gen 15 3.8/4.1 234000G010 234000G015 ...(KA'S) =
n m
w -
- v-m-
p,. -..mr--m+- y wm-7 y -.-w- --'"--9 --
s' a TEST CROSS REFERENCE PAGE
ESI!ON VALUE REFERENCE ______ ______ __________ 05.01 3.00 CXG0004122 05.02 2.50 CXG0004123 05.03 2.50 CXG0004131 '05.04 2.00 CXG0004135 05.05-1.50 CXG0004136 05.06 3.00 CXG0004138 05.07 2.00 CXG0004140 05.08 1.50 CXG0004141 05.09 2.50 CXG0004142 - 05.10 1.50 CXG0004177 05.11 1.50 CXG0004179 03.12 1.50 CXG0004183 ______ 25.00 06.01 2.00 CXG0004143 06.02 3.00 CXG0004144 06.03 3.00 CXG0004145 06.04 2.00 CXG0004146-06.05 3.00 CXG0004147 "06.06 2.00 CXG0004148 06.07 2.00 CXG0004149 06.08 3.00 CXG0004150 06.09 2.00 CXG0004152 06.10 1.00
- XG0004181 06.11 2.00 CXG0004184
______ 25.00 07.01 3.00 CXG0004153 07.02 2.00 CXG0004154 a07.03 2.00 CXG0004155 07.04 3.00 CXG0004157 07.03 2.50 CXG0004158 07.06 2.25 CXG0004159 07.07 2.00 CXG0004160
- 07.08 2.00 CXG0004161 07.09 1.50 CXG0084162 a07.10 2.25 CXG0004163 07.11 2.50
'CXG0004180 ______ 25.00 08.01-3.00 CXG0004165 08.02 1.50 CXG0004166 08.03 1.00 CIG0004167 08.04 2.00 CXG0004168 08.03 3.00 CXG0004169 08.06 3.00 CXG0004170 08.07 2.50 CXG0004172 )
- -. , . TEST CROSS REFERENCE PACE
, [E S'T I ON VALUE REFERENCE p_.____ ______ __________ '08.00 2.50 CXG0004173 08.09 2.00 CXG0004174 08.10 2.00 CXG0004176 08.11 2.50 CX80004182 ______ 25.00 ______ ______ 100.00 DOCKET-NO 293
. -- YOLbAtk
0 . BOSTON EDISDN Executive Offices 800 Boylston Street Boston, Massachusetts 02199 Ralph G. Bird Senior Vice President - Nuclear December 11, 1987 U.S. Nuclear Regulatory Commission Region I 631 Park Avenue King of Prussia, PA 19406 Attention: Mr. Allen Howe Subject: Licensing Examination Hritten Comments
Dear Mr. Howe:
In accordance with the criteria established in NUREG-1021, Section ES-201, Part "H," our Operator Training Staff has prepared the enclosed written comments for your review and consideration.
As the written examination was administered to our license candidates on December 8,1987, the five working day limit for providing feedback on the written examination is satisfied.
The professional manner in which the examination administration activities of this week have been conducted is appreciated.
Any questions on the attached examination comments, should be directed to Mr. Harrison R. Balfour, at (617) 746-6578.
/' j . R. G.
ird ARS/lo Enclosure cc: E. J. Ziemianski H. R. Balfour ! . _.
.-. . . .. . . -.
18760/5 QUESTION 5.04: (2.00 Points) - Indicate whether the following statements concerning fission product poison behavior are TRUE or FALSE for your reactor.
IF FALSE, change the statement so that it is correct.
(Consider each statement separately) a. Equilibrium xenon concentration at 50% power is approximately half of the equilibrium concentration at 100% power (0.5) b. Equilibrium samarium concentration is the same for all power levels.
(0.5) c. Both xenon and samarium concentrations increase immediately after a reactor shutdown from high powers.
(assume equili-brium had been reached before shutdown) (0.5) d. Xenon-135 decays with a half-life of about 4.5 hours while samarium-149 is stable.
(0.5) ANSHER: a. False (0.25) Equilibrium level at 50% is approximately 2/3 the equilibrium level at 100% power (0.25) b. True (0.5) c. True (0.5) d. False (0.25) Xenon has a half-life of about 9.2 hours (0.25) REFERENCE: G. E. Reactor Theory Hanual, pages 6-8, 6-15, 6-11 and 6-13 Enabling objective 2.1, 2.5, 2.6, 3.5, 3.6, Chapter 6 K/A's.. 292006 K1.03 2.9/2.9 292006 K1.13 2.6/2.6 292006 K1.04 2.9/2.9 292006 K1.07 3.2/3.2 292006K103 292006K104 292006K107 292006K113 ...(KA's) COMMENTS: / a. Request accepting an answer that shows that the student understands xenon concentration at 50% power is greater than 50% of the concentration at 100% power.
Answer key states "approximately 2/3".
Exact values are not provided in the training material.
l REFERENCE: Reactor Theory Licensed Operator Student Guide, pages 6-10a and 6-11a j PNPS Reactor Theory HanJbook, pp. 28 and 29 . .. _ _. _
. - -... - 18760/7 QUESTION 5.06: (3.00 Points) - Hill the following cause core reactivity to increase or decrease? Briefly explain for each why reactivity will change.
, Assume reactor is at full power.
a. Loss of a feedwater heater (1.0) b. Sudden increase in reactor pressure (prior to a reactor scram) (1.0) c. Build-up of corrosion products of the fuel pins.
(1.0) ANSWER: a. Reactivity increases (0.25) the cooler water entering reactor causes increased neutron moderation and contact time with the fuel shifting the boiling boundary up in the core (0.75) b. reactivity increases (0.25) due to collapse of voids, increase in moderator density, with more thermalization (0.75) c. reactivity decreases (0.25) due to higher fuel temperature more parasitic absorptions (doppler effect) occur in the resonance region (0.75) REFERENCE: G. E. Heat Transfer and Fluid Flow Manual, pages 9-16, 9-19 and 9-33.
Enabling objective 3.1, 3.6 4.1, 4.2, S.1 and 5.3, Chapter 9 K/A's.. 292004 Kl.01 3.2/3.2 292004K101 ...(KA's) COMMENTS: a. Request that "contact time with the fuel shifting the boiling boundary up in the core" be put in parentheses, and not required for full credit.
Explained in the attached reference (Paragraph 2) is a typical discussion of the effects for a change in inlet subcooling. Although the boiling boundary does shift, it is more of a secondary heat transfer effect than a means to explain the reactor physics associated with a change in moderator temperature, b. NO COMMENT c. NO COMMENT REFERENCE: Reactor Theory L0 Student Guide, page 4-47.
.o..,/10 00ESTION 5.DS: (2.50 Points) - You are performing a reactor startup and the reactor is critical at a power level of 2 KH.
The reactor operator withdraws a cortrol rod 2 notches and reactor power increases with a doubling time of 90 seconds.
Answer the following showing all work: a. What is the reactor period calculated in seconds? (0.75) b. Assuming no further operator action, how long will the reactor power level take to reach the point of adding heat (POAH - 3.3 MH) (0.75) c. When the POAH is reached, how much will moderator temperature change before the temperature coefficient (ALPHA H00ERATOR - -lx10E-4 delta K/K/F) causes power to stop increasing.
(1.0) ANSHER: a. Period - Doubling Time x 1.443 - 90 sec. x 1.443 - - 130 seconds (0.75) b. P(f) - P(i) eE (t/T) 3.3 x 10E6 / 2000 - eE (t/T) T - 963 sec. or 16.03 min.
(0.75) c. reactivity .0072/ (.1 x 130) +1) .0005 delta K/K delta T - alpha T x reactivity added - -1 x 10E-4 x.0005 delta K/K/F - 5 F increase in temperature (1.0) (note to grader - part c is to be graded on methodology) REFERENCE: G. E. Reactor Theory Manual, Chapter 5 Objective 2.7, Chapter 4 Objective 2.5, Chapter 3 Objective 3.5, 3.6, 3.7 K/A's.. 292003 Kl.05 3.7/3.7 292003 K1.08 2.7/2.8 r 292003 Kl.09 2.5/2.6 292003K105 292003K108 292003K109 ...(KA's) Q)ttiDilS: 5.09 c. Request that you consider alternate methodologies for obtaining answer. Although it was expected by the examiner for candidate to use information derived in part "b", the candidate may use a method which utilizes the data given in the question rather than relying on the correctness of his answer to the previous part.
REFERENCE: Reactor Theory L0 Student Guide, page 5-19 & 5-20a.
. . - 18760/14 . OUESTION 6.01: (2 00 Points) - a. State TH0 (2) conditions that will cause an auto isolation of RCIC.
(Include setpoints) (20 0.5) D. List FOUR RCIC turbine trips and include setpoints if applicable.
(4 0 0.25 ea) ANSHER: a. Reactor pressure low, 50-100 psig Hi steam supply delta p. 300% for > 3 sec Hi turbine area or steam line temperature, 150-200 F in torus, RCIC room RCIC valve room (any 2, 0.25 for trip, 0.25 for setpoint) b. Turbine overspeed 0 125% Low suction pressure; 25" vacuum Hi turbine exhaust pressure, 25 psig Auto isolation signal Manual pushbutton at the 904 panel Manual trip at the turbine (any 40 0.25 ea) REFERENCE: PNPS Student Guide 09-04, Enabling Objective 11, 13 K/A's.. 217000 K4.04 3.0/3.1 217000K404 ...(KA's) , ! ! COMENTS: a. Request that the "for greater than 3 sec." portica of the answer e placed in parentheses, and not required for full r.redit.
The setpoint for the RCIC high steam supply flow isolation is 5 M of rated steam flow.
r b. Request that the trip setpoint of 15 inches of vacuum be accepted for full-credit.
The correct setpoint for the low pump suction pressure trip for the RCIC Turbine is 15 inches vacuum.
The examination answer key shows a trip setpoint of 25 inches of vacuum.
l l l l REFERENCE: PNPS RCIC Student Guide, page LG-ll.
18760/18 OUESTION 6.05: (3.00 Points) - , a. Describe SBGT system response to a loss of heating element during operation of the system.
(1.5) b. State the reasons for changing the deluge fire system from automatic operation to manual.
(1.5) ANSHER: a. If a heater trips or burns out a current sensing relay will trip the SBGT system blower for that train.
(0.5) If it were allowed to operate without heaters, the moisture content would increase and decrease the efficiency of the charcoal filters. (1.0) b. Manual operation of the deluge system eliminates the possibility of an automatic initiation of the deluge system after a design based accident (which could wash radio-iodines into the TB floor drains) (1.5) REFERENCE: PNPS Student Guide 08-03 Learning Objective 11B, 140, 11D PNPS Misc Matis PDC-86-70 Learning Objectives 1, 5 K/A's.. 261000 K4.02 2.6/2.8 261000 K4.03 2.5/2.7 261000 K4.06 2.4/2.6 261000 K5.01 2.3/2.6 261000K402 261000K403 261000K406 261000K501 ...(KA's) COMENIS: b. Request that the words "after a design based accident" be placed in parentheses and not required for full credit.
l This question asks the examinee to state the reason (s) for changing the SBGT Deluge Fire System from automatic operation to a manually operated system.
The reason is to "prevent inadvertent spraying of
I the charcoal filters", if a malfunction in the automatic system were to occur.
E l
REFERENCE: PNPS Trainee Handout for PDC package # PDC-86-70, "Description of Change" page.
_ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ ic/bv... QUESTION 6.0j: (3.00 Points) - a. You have completed verifying an isolation of the reactor building when a report comes in that instrument air to the isolation dampers will be lost.
What position will the isolation dampers end up and why? (1.0) b. List FOUR signals that will cause a reactor building isolation and an initiation of Standby Gas.
(Give setpoints were applica' ole) (2.0) ANSHER: a. The valves will remain shut (0.25) because they fail shut upon loss of instrument air.
(0.75) b. Reactor Low Level (+9") High Drywell Pressure (+2.5 psig) High Refuel Floor Rad Levels (> 16 mr/hr) Four Downscale Refuel Floor Exhaust Duct Radiation Monitors Two Downscale combined with One Upscale Refuel Floor Exhaust Duct Rod Monitors (4 @ 0.5 ea) REFERENCE: Student Guide 08-05 Plant Ventilation, Enabling Objectives, 5, 6, 7 K/A's.. 288000 Kl.02 3.4/3.4 288000 K4.01 3.7/3.9 288000 K4.02 3.7/3.8 288000 K6.03 2.7/2.7 288000K102 288000K401 288000K402 2888000K603 ...(KA's) COMMENTS: a. NO COMMENT b. Request that the following be included in the answers that are listed as acceptable answers.
/ Loss of 120 vac panel #Y-2 e REFERENCE: PNPS Procedure #5.3.6, "Loss of Vital A.C. (Y-2), page #5.3.6-2.
t
\\ -18760/22 00ESTION 6.09: (2.00 Points) - The reactor is at 70% power when the A Flow Converter fails such that its output is DOHNSCALE. State ALL trips that will occur and all causes for each trip. A center rod was selected at the time of the failure.
(2.0) ANSHER: R00 BLOCKS - flow comparator mismatch (> 10%) (0.5) - flow biased trip (power >.58 H + 50) (0.5) - RBH flow bias trip (0.5) HALF SCRAM CHANNEL A - flow bias (power >.58H + 62 (0.5) REFERENCE: PNPS Student Guide 07-01 NHS pp LG-12, 13, 14, 15 K/A's.. 212000 K6.02 3.7/3.9 215005 K1.10 3.3/3.4 215002 Kl.01 2.9/3.0 215005 Kl.16 3.3/3.4 215005 K6.07 3.2/3.3 212000K602 215000K110 215005Kil6 215005K607 ...(KA's) COMMENTS: For the rod blocks which are caused as a result of this malfunction, request that a "Flow Converter Inoperable Trip" be added as an additional acceptable response. A "Flow Converter Inop" trip will occur and initiate a rod block if either Flow Converter Unit fails downscale, is unplugged or has its function switch taken out of OPERATE.
REFERENCE: PHPS APRM Reference Text, page #APRM-4-2 r _ _ _ _ _.
18760/25 OUESTION 7.01: (3.00 Points)
Using the attached curves from E0P-03, answer the following: a. What is the minimum torus water level for RPV pressure of 900 psig and torus temperature of 165 degrees F? (1.0) b. What is the maximum RPV pressure with a torus water level of 7 feet - 7 inches at a temperature of 154 degrees F? (1.0) c. If the torus level COULO NOT be maintained above the Heat Capacity Level Limit, WHAT two immediate actions must be taken and HHY? (1.0) ANSHER: a. 103 inches (8 feet 7 inches) (1.0) b. 800 psig (1.0) c. If suppression pool level cannot be maintained, manually scram the reactor (0.33), and depressurize the RPV using ADS (E0P-07) (0.33).
This is to ensure that condensation of the steam in the vessel in the event of a LOCA can be assured.
(0.33) REFERENCE: PNPS Technical Bases of E0P's Primary Containment Control - Sect. III.D K/A's.. 2230001G014 3.8/3.7 223001G014 ...(KA's) C0WENTS: a. Request that the answer key have a tolerance band of i 5 inches to allow for interpolation.
b. Request that the answer key have a tolerance band of i 50 psig to allow for interpolation.
>, c. Request that part C be deleted and points for C be redistributed to parts A and B.
There are no immediate actions for our E0Ps.
Students are not required to know steps from memory.
REFERENCE: PNPS E0P Student Guide, SG-2 Objective 3
18760/2c OUESTION 7.02: (2.00 Points) - OTHER THAN ALARMS, what FOUR symptoms are indicative of a small !eak (drywell press. < 2 psig) inside the primary containment? (2.0) ANSHER: Excessive sump pump operation (due to increase leakage to the drywell equipment drain sump or drywell floor drain sump.)
(0.5) Abrupt change in drywell humidity (as indicated on panel C85 recorders) (0.5) High radiation detected on drywell leakage detection system (panel C19).
(0.5) Significant pressure changes as recorded on drywell pressure recorders.
(0.5) REFERENCE: PNPS Procedure 2.4.14 Leakage Inside the Primary Containment PNPS Tech Specs 3.6.C.1 COMMENTS: Procedure 2.4.14 also lists "Significant increase in drywell temperature" as a symptom.
Request this symptom to be acceptable as an additional answer.
BEff.HENCE: PNPS Procedure 2.4.14 f I l l
18760/28 DUESTION 7.04: (3.00 Points) - If operating in shutdown cooling and all recirculation flow is lost, PNPS Procedure 2.4.24 (Reactor Vessel Cold Hater Stratification) states to monitor instrumentation for verification of cold water stratification, a. What are the indications of stratification? (0.5) b. Other than restoring shutdown cooling or recirculation flow stratification can be minimized by affecting the flow in two systems. WHAT are these systems and H0H ARE THEIR FLOHS ALTERED to minimize stratification (be specific)? (1.5) c. One concern when recirc flow is lost is the possibility of losing natural circulation. WHAT can be done to ensure natural circulation will continue? (1.0) ANSHER: a. A large differential temperature between the Reactor vessel bottom and elsewhere in the vessel (as read on recorders TR263-105 and TR263-104).
(0.5) b. RHCU and CRD (0.5) Throttle down on RHCU cuction (HO-1201-85) from the recirculation loop (keep RHCU suct. press > 20 psig) Reduce CRD flow (2 @ 0.5 ea) c. Raising vessel water level to +46 inches (instrument zero) ensures natural circulation will occur (1.0) REFERENCE: ' PNPS Procedure 2.4.24 Cold Hater Stratification PNPS Procedure 2.4.25 Loss of Shutdown Cooling K/A's.. 295021 Kl.02 3.3/3.4 295021 K1.04 3.6/3.7 295021 K3.01 3.3/3.4 295021 Kl.01 3.4/3.4 295021A101 295021K102 295021K104 295021K301 ...(KA's) r l l COMMENTS: a. NO COMMENT b. NO COMMENT i c. Request level +46 be put in parentheses as long as the discussion ' includes raising level high enough to assure flow through the separators.
REFERENCE: PNPS Procedure 2.4.24 Cold Hater Stratification PHPS Procedure 2.4.25 Loss of Shutdown Cooling !
.60/31 OUESTION 7.07 (2.00 Poin*s) - a. Jet pump failure per PNPS 2.4.23 "Jet Pump Flow Failure" can be indicated when there is a sudden change in (1) (0,5) , without a corresponding change in (1) (0.5) . b. According to Tech Spec 3.6.E and 3.6 F. if a jet pump is considered inoperable (failed) the reactor must be brought to ' cold shutdown within 24 hours. Why is jet pump failure such a serious problem.
(1.0) ANSHER: a.1. core differential pressure, recirc flow, OR jet pump flow indication (0NLY ONE REQUIRED) (0.5) 2. recirc pump speed (0.5) b. Loss of a jet pump could preclude the capability to maintain 2/3 core coverage during a LOCA.
(1.0) REFERENCE: PNPS 2.4.24 Jet Pump Flow Failure PNPS Technical Specifications 3.6 E & F, and Bases K/A's.. 202001 A2.01 3.4/3.9 202001 Gen 6 3.0/4.1 202001 K5.02 3.1/3.2 2020001A201 202001G006 202001K502 ...(KA's) COMMENTS: a. NO COMMENT b. Request that consideration be given to including "increase in blowdown area" as part of the answer.
PNPS Technical Specification bases state that failure of a jet pump will increase the cross-sectional flow area for blowdown following a DBA LOCA.
r REFERENCE: Tech. Spec. Bases 3.6.E and 4.6.E, page 147.
.-. . . - .. '18760/35 OUESTION 7.11: (2.50 Points)
Fill in the following blanks with the maximum permissable value: a. The average rate of change for a normal heatup or cooldown shall not exceed (1) F when averaged over one hour.
(0.5) b. When attempting to open an MSIV, differential pressure across the valve shall not exceed (2) psid.
(0.5) c. Do not start an idle relationship pump if the differential temperature between the idle loop and the operating loop exceeds (3) F OR if the differential temperature between the vessel steam dome and the bottom head drain is greater than (4) F.
(1.0) d. The main condenser vacuum pump must be secured prior to the reactor exceeding (5) % thermal power.
(0.5) ANSHER: a. 100 b. 200 (each c. 50; 145 correct answer d. 4 0.5) , REFERENCE: PNPS 2.1.9 Reactor Recirculation Pump Operation, Precaution A PNPS 2.1.3 Hot Standby Haneuvers Startup with MSIV's Closed PNPS Tech Spec 3.6.A PNPS 2.2.92 MS Lino Isolation and Turbine Bypass Valve PNPS 2.2.93 Main Condenser Vacuum System K/A's.. 239001 K4.09 3.3/3.3 239001 A4.01 4.2/4.0 239001 Gen 5 3.2/4.2 239001A401 239001G005 239001K409 ...(KA's) COMMENTS: J' b. Request that consideration be given to accepting 50 psid as the actual differential pressure limit across MSIVs prior to opening of the valves.
Precautions in PNPS Procedure 2.2.92 caution the operator to not attempt to open MSIV with a differential pressure of > 200 psid across it.
This is a design parameter and prevents valve damage.
Step 7 of the procedure for opening MSIVs with the reactor pressurized states a conservative operational differential pressure limit of 50 psid across the HSIVs prior to opening of the valves.
REFERENCE: Procedure 2.2.92, pages 7 and 8 -- .. - _ _ . -.
'18760/36 QUESTION 8.01: (3.00 Points)
The reactor is operating at a high power level when a TURBINE TRIP occurs causing the reactor to scram on HIGH FLUX. Answer the following: a. Has a violation of SAFETY LIMITS occurred? If so why; if not, why not? (1.5) b. What immediate actions are you as Watch Engineer required to take? (2 required) (1.5) ANSWER: a. Yes, a violation has occurred.
(0.5) A Safety Limit is assumed to have been exceeded if a scram is initiate by means other than the primary source signal (in this case, the control valve fast closure). (1.0) b. Shutdown the reactor Notify the NRC within one hour (2 0 0.75 ea) REFERENCE: PNPS Technical Specification 6.7 and 1.1.C SRO Augmentation 01-04 Objective 8 K/A's.. 295006 Gen 11 4.5 295006 Gen 4 4.2 295006 Gen 1 4.5 295006 Gen 1 4.1 295006G001 295006G002 295006G004 295006G011 ...(K/A's) COMMENTS: i t a. NO COMMENT b. Request that any two for the four actions listed in Tech. Specs. be accepted, as an answer to part b.
T REFERENCE: l Tech Specs. Section 6.7, page 217 ! l
. _... '6,....v OUESTION 8.05: (3.00 Points) - a. According to PNPS 1.4.5 "PNPS Tagging Procedure" who (by title) is authorized to approve tagout sheets so that pro-tective tags can be put in place? (0,5) b. For what two types of plant systems is an independent second verification of tag placement required? (0.5) c. TRUE or FALSE: (if FALSE, correct so that it is TRUE) 1. Haster Damage tags are used to isolate major components (e.g., pumps, compressors, etc.) where Red Tags are used for minor components (e.g., switches, valves, fuses, etc.)
(0.5) 2. Caution tags are suited for components like the SLC control switch in the main control room when one of SLC pumps is out for maintenance and the other is required for operation.
(0.5) d. What must be done to ensure tags temporarily removed from a maintenance boundary (isolation) are documented and who (by title) is responsible for making this documentation? (1.0) MSBER: a. Hatch Engineer or Operating Supervisor (on-duty) has authorization for issuance.
(0.5) b. An independent verification is required on Nuclear Safety Systems and Auxiliary Systems that support Nuclear . Safety Systems.
(2 0 j 0.5 ea) c. 1. FALSE (0,1); Master tags are used to keep control room and management personnel cognizant that maintenance is being performed on a system (0.2).
Red Tags are attached to a switch or valve to prohibit the operation of the device, to protect personnel from injury and equipment from damage.
(0.2) 2. TRUE (0.5) d. The maintenance supervisor (0.5) must provide a written reason for the tags removal in the comment block on the r tagout sheet for the tags to be removed.
(0.5) REFERENCE: PNPS 1.4.5 PNPS Tagging Procedures K/A's.. 294001 K1.02 3.9/4.5 294001K102 ...(KA's) ! l l l
^ - - .. 18760/41 COMENTS:
a. Request that either Hatch Engineer or Operating Supervisor be considered for full credit.
b. NO COMMENT c. NO COMMENT d. NO COMMENT REFERENCE: Procedure 1.4.5, page 9
> ,. , . - - -.. . - .. .... - - .. - . -. -. ...... -. -. - - - .. - -. - - - _ - - -. .- - - _
18760/44 00ESTION 8.08: (2.50 Points)
Concerning control rods: a. Tech Specs states that a maximum of eight control rods can be inoperable while at power. What is the basis, per T.S. BASES 3.3 and 3.4, for this number? (1.5) b. TRUE or FALSE; A fully inserted control rod that has been electrically disarmed is considered inoperable.
(1.0) ANSWER: a. (Particularly late in life, shutdown margin can be assured with more than 8 inoperable (partially withdrawn) rods); a core experiencing more than 8 inoperable rods is indicative of a generic rod drive problem and the reactor is shutdown.
(1.5) b. False (1.0) REFERENCE: PNPS Technical Specifications 3.3.A and C. Bases 3.3 and 3.4, Definition E SRO Augmentation 01-01, Enabling Objective 3.d 01-03, Enabling Objective 4 K/A's.. 201003 Gen 63. 3/3.9 201003 Gen 5 2.8/3.7 COMMENTS: a. Request that you consider for full credit a discussion on collet housing failures as the basis for 8 rods INOP.
b. NO COMMENT r REFERENCE: T. S. Bases, 3.3 and 4.3, page 88 . __ _ __ . - _ _ _.
18760/ a 00ESTION 8.09: (2.00 Points) - According to procedure PNPS 1.3.6 "Adherence to Technical Specifications" with proper consideration and authorization, licensed operators can knowingly deviate from License Conditions or Tech Specs.
' a. To deviate from tech specs without a "Tech Spec Clarift-cation", personnel considering protective action must consider two criteria first.
What are the two criteria.
(1.0) b. Who (by title) may approve this action, as a minimum, and what additional action must be taken, if time permits, prior to initiating the protective action.
(1,0) ANSHER: a. It is necessary to protect the health and safety of the public No action consistent with the license or Tech Specs that can provide adequate or equivalent protection is immediately apparent.
(29 0.5 ea) b. As a minimum, a licensed SRO must approve the action (0,5) Notification by phone to the NRC PRIOR to the action taking place IF TIME PERMITS (0.5) REFERENCE: PNPS 1.3.6 Adherence to Tech Specs PNPS Tech Specs Section 6 294001K116 ...(KA's) COMENTS: ! a. NO COMMENT b. Request that you consider for full credit the Hatch Engineer.
Where the question is worded "Who By Title" student may be led to answer the position (Licensed SRO) as the Hatch Engineer.
l ! .. ._.
. - -. _. . - -. . . _ _.
18720/4 OUESTION 1.03: (3.00) - For the following changes in plant parameters HILL control rod worth INCREASE, DECREASE or NOT BE AFFECTED? Briefly EXPLAIN HHY? a. An increase in moderator temperature.
(1.00) b. An increase in void content.
(1.00) c. An increase in fuel temperature.
(1.00) ANSWER: a. As moderator temperature increases the leakage of thermal neutrons from the fuel bundles into the control rod regions increases.
(0.50) Thus rod worth increases.
(0.50) b. Moderator density decreases resulting in more fast neutrons and fewer thermal neutrons leaving the bundle.
(0.50) Since control rods are thermal absorbers, overall control rod worth decreases.
(0.50) c. Not affected.
(0.50) Since fuel temperature affects primarily fast neutrons, which are resonantly captured, and control rods are thermal neutron absorbers, fuel temperature and rod worth are essentially independent of each other.
(0.50) REFERENCE: Pilgrim: Reactor Theory, pp. 5-12 through 5-14.
KA 292005X019 (2.5/2.6) COMMENTS: a. NO COMMENT b. Consider as alternate answers 1. An answer that relates an increase in void content with additional suppression of local flux (worth proportional to local flux / average flux squared) which results in lower rod worth.
2. An answer that addresses increased resonance capture resulting in lower thermal flux which reduces rod worth, c. NO COMMENT REFERENCE: Reactor Theory Student Guide, pp. 5-13a and 5-29 PNPS Reactor Theory Handbook, pp. 20 and 21
. v i 8v.. QUISTION 1.04: (2.00) - The reactor has been operating at 95 percent power for several days.
An operator RAPIOLY reduces reactor power to 60 percent by reducing the speed of the recirculation pumps.
During the next 2-3 MINUTES the operator notices that the reactor power slowly increases to 63 percent (with no operator action).
EXPLAIN the cause of the power increase.
(2.00) ANSWER: The reactor is now producing less steam to go to the turbine.
There will be less extraction steam and reheater drain steam going to the feedwater heater.
(1.00) Therefore, less feedwater heating will occur resulting in colder feedwater entering the vessel (0.50) which will cause reactor power to increase (about 3 percent) from the positive reactivity addition (alpha m).
(0.50) REFERENCE: Pilgrim: HT&FF (GE), pp. 5-48.
Pilgrim: Reactor Theory (GE), pp. 7-18 through 7-20.
KA 29008K120 (3.3/3.4) 292008K121 (2.9/3.0) 293005K105 (2.7/2.8) COMMENTS: Request "and reheater drain steam" be put in parentheses.
Reheaters are not used at Pilgrim Station, REFERENCE: HT&FF Student Guide pp. 9-36 and 9-37 9' c
- . 18780/10 00ESTION 1.09: (2.00)
, HHAT are four (4) plant or component design features or operational limitations which ensure adequate net positive suction head for the reactor recirculation pumps? (2.00) ANSWER: Any four (4) (0.50) each, 2.00 maximum 1. Pumps are located below normal water level.
2. Minimum speed interlo:k is enforced when feedwater flow is less than 20 percent.
3. At high power operation, feedwater subcooling provides increased net positive suction head.
4. Recirculation pump runback on low reactor vessel water level.
5. Recirculatic, pump trips (or will not start) if suction valve is not full ()en (90 percent).
6. Recirculation pump runback (speed limit) if discharge valve is not full open (90 percent).
7. Recirculation pump trip on low-low water level.
8. Recirculation pump trip if discharge valve control switch is placed in CLOSE REFERENCE: Pilgrim: Student Guide 0-RO-02-06-02, pp. LG-3 and LG-4.
Pilgrim: Systems Reference Texts, Reactor Recirculation, pp. 21, 22 and 35.
KA 20200lK402 (3.1/3.2) 291004K106 (3.3/3.3) 293006K110 (2.7/2.8) COMMENTS: J' Request that the following additional alternate answers be considered.
, - Procedural requirement to open the discharge valve within 10 seconds after a pump start.
- Procedural requirement for water level above 35" prior to pump start.
j - Procedural requirement to maintain a normal water level band.
l ' REFERENCE: Procedures 2.2.82 and 2.2.84 .
18780/12 QUESTION 1.11: (2.00) - a. Most condensers are designed with excess condensing capability; that is, the condensed liquid leaves the condenser hotwell several degrees below the saturation temperature.
iiOH would PLANT EFFICIENCY be affected (INCREASE, DECREASE or NOT AFFECTED) if the temperature of the circulating water was greatly DECREASE 0? EXPLAIN your answer.
(1.00) b. If the main condenser was absolutely air tight, HOULO there be any need for the air ejectors? Explain HHY.
(1.00) ANSWER: a. (Although turbine efficiency would increase) overall plant efficiency would decrease (0.25) because the heat rejected to the recirculating water must be added to the feedwater by the reactor.
(0.75) b. Air ejectors would still be needed in order to maintain condenser vacuum because air in-leakage is not the only source of noncon-densibles to the main condenser.
(0.50) Other NC include radiolytic 02 and H2 and fission product gases.
(0.50) REFERENCE: Pilgrim: HT&FF (GE), pp. 5-12 through 5-14.
KA 293004Kil2 (2.9/3.1) 291006Kil8 (2.8/2.9) COMMENTS: a. NO COMMENT b. Request that "and fission product gases" be put in parentheses.
The trainees were taught that the volumetric contribution of fission product gases is considered negligible, f REFERENCE: FSAR Section 9.4 l l -- .. --
. - _.. _. _.. _ - 18780/13 . ' QUESTION 2.01: (2.00) For the Rod Block Monitor (RBM), PROVIDE answers to the following questions: a. WHAT adverse condition is the system designed to prevent? (1.00) b. When the Heter Function Switch on the Back Panel 937 Meter Section is on the "Count" position. HHAT are the "units" of the indication on the meter and HHAT can be calculated by utilizing the indicated value? (1.00) ANSHER: a. Local fuel damage (by generating a rod withdrawal block).
(1.00) b. Units - volts (0.50), number of operable LPRM inputs can be calculated (by using 1 volt per operable input).
(0.50) REFERENCE: Pilgrim: Systems Reference Texts, RBM, pp. I and 11.
KA 215002SG04 (3.3/3.4) 215002K102 (3.2/3.1) 215002A402 (2.9/2.9) COMMENTS: a. Consider alternate answer to crevent exceeding MCPR b. Request the full credit be given for the last part of the answer in part B.
PNPS Reference Text are used to develop training material for the Nuclear Training Department.
Students are not held accountable for all material in the Reference Text.
They are held accountable for the objectives in their specific course.
This part of the answer is not listed as an objective.
Reactor Operators at PNPS do not perform Rod Block Monitor Surveillances.
T This is done by the I&C Group.
REFEREECE: Tech. Specs. p. 205C-4 and Tech. Specs. 91 0-RO-02-07-01, p. 3
18780/14
QUESTION 2.02: (3.00) For each of the following situations, DETERMINE whether or not the activity can occur.
If the activity can NOT occur, HHAT must change to allow it to occur.
a. Refuel bridge is over the reactor vessel and in motion toward the fuel pool with the fuel grapple loaded. All rods are inserted.
The reactor mode switch position is changed from REFUEL to START-UP. Hill the bridge continue to move? (1.00) b. Refuel platform is over the vessel.
The frame mounted hoist is loaded. One rod is at position 30.
CAN the load on the hoist be lowered into the vessel.
(1.00) c. Refuel platform is over the vessel with the mode switch in REFUEL.
The grapple is fully lowered and unloaded.
CAN a control rod be withdrawn? (1.00) ANSHER: a. Yes (1.00) b. No (0.50); must insert the rod.
(0.50) c. No (0.50); must raise the grapple fully or move refuel platform away from the core.
(0.50) REFERENCE: Pilgrim: Systems Reference Texts, Refueling, pp.7, 28 and 29.
KA 23400K502 (3.1/3.7) 234000A302 (3.1/3.7) COMMENTS: a. NO COMMENT b. NO COHHENT c. Request that for the second part of the question worth (0.50) the points be awarded for either "must raise the grapple fully" or "move refuel platform away from the core".
REFERENCE: PNPS Student Guide, Fuel Handling Equipment, p. LG-5 i _. _ _. _..
- _ __ leidO/16 ' OUESTION 2.04: (2.50) Concerning the drywell leak detection system: a. A drywell equipment drain sump high level is annunciated in the control room.
If level continued to increase, HHAT three (3) other signals or actions occur as a direct result of a HIGH LEVEL and HIGH-HIGH LEVEL? (1.50) b. The drywell equipment drain sumps and the drywell H2/02 Honitoring System isolate on a Croup II PCIS signal.
DESCRIBE HHAT operator action can be taken, if any, to place the H2/02 monitor back in service before the Group II PCIS signal is cleared.
(1.00) ANSHER: a. 1. (High level switch provides) a start permissive to one sump (0.50 pump (when pump is in AUTO mode), each) 2. (High high level switch provides) a start permissive for second sump pump.
3. (High level switch) starts a 30-minute adjustable timer.
b. Place the H2/02 valve control switches in CLOSE (0.33) place the override keylock switches in OVERRIDE (0.34) and place the H2/02 valve control switches back in OPEN.
(0.33) REFERENCE: Pilgrim: Systems Reference Texts, DLD, pp. 3 and 13.
KA 223001K104 (3.2/3.3) 223001K110 (3.4/3.6) COMMENTS: a. Request that the first two answers in the answer key be accepted for full credit.
The drywell equipment drain sump pump 30 minute timer if started by depressing the pump start switch (es).
Request consideration be given to "alarm in radwaste control room (C-20) as an alternate answer, b. NO COMMENT REFERENCE: Primary Containment Student Guide 0-RO-02-08-01, page LG-38.
Drywell Leak Detection Reference Text, page DLD-11.
18780/18 ' OUESTION 2.06: (3.00) While the reactor is operating at 100 percent power, a complete loss of essential instrument air occurs. Assuming no operator action.
DESCRIBE H0H and HHY the following parameters will change prior to a reactor scram.
(3.00) a. Main condenser vacuum b. RBCCH temperature c. CRD cooling water flow d. FH flow rate to the reactor vessel e. FH temperature f. Indicated SLC tank level ANSHER: a. Decrease as the SJAE supply valves close (or because off-gas (0.50 isolates).
each) b. Decrease because RBCCH HX bypass valve closer, providing maximum cooling to RBCCH.
c. Decreases as CRD FCV's close.
d. Decreases as FH regulating valves fail as-is and condensate pump and FH pump recirculation valves fail open.
e. Decreases because bleeder trip valves fail shut (and spill valves fail open) thus eliminating all FH heating.
f. Tank level indication goes to zero since level in:trument requires air to operate (but no change in actual tank level).
REFERENCE: Pilgrim: Systems Student Guide, Instrument Air, pp. 12 and 13.
KA 295019AK01 (3.8/3.9) 295019AK102 (2.9/3.0) 295019AK103 (3.2/3.3) 295019AK105 (3.4/3.4) 295019AK106 (2.8/2.9) 295019AKl15 (2.3/2.6) o COMMENTS: a. Request that you accept the following as an alternate answer for synonymous terminology "Decrease as tne vapor valves close."
b. Request that you accept the following as an alternate answer for synonymous terminology "Decrease because RBCCH temperature control valve Closes."
REFERENCE: Main Condenser Vacuum and Air Removal SG, 0-RO-02-04-03, pp. LG-3 and LG-4.
Main Condenser Vacuum and Air Removal Reference Text, figure 2.
RBCCH SG, 0-RO-02-02-06, page LG-11.
18780/19 QUESTION 2.07: (2.50) - a. STATE the normal and alternate power supplies to the Reactor Protection System (RPS) bus.
(1.00) b. HHICH of the following components would be directly affected by a manual transfer of RPS Bus A from its normal to alternate power supply? (1.50) 1. IRH A 2. APRH B 3. Reactor building ventilation radiation monitors 4. Off-gas system radiation monitors 5. MSIV 6. Main steam line radiation monitors ANSWER: a. Normal - RPS HG sets (0.50) (or HCC B-23 and MCC B-22) Alternate - 480V HCC B-10 (0.50) b. 3. (Reactor building ventilation radiation monitors) 4. (Off-gas system radiation monitors) 6. (Main steam line radiation monitors) (0.50) each, (-0.50) for incorrect answers.
REFERENCE: Pilgrim: Systems Reference Text, Reactor Protection System, pp. 9-3 and 36-2.
KA 212000A202 (3.7/3.9) 212002K201 (3.2/3.3) COMMENTS: r a. Request that grading consideration be given to responses in terms of load centers (i.e., B3 feeds MCCB23, B4 feeds MCCB22, and B6 feeds HCCB10).
Question does not specifically ask for power supplies in terms of HCCs; therefore, some candidates may respond with load center designations, b. Request that Reactor Building Ventilation radiation monitor be deleted as part of the answer to 2,07 part b., with the 1.50 points being equally redistributed over the remaining two answers.
The power supply to reactor building ventilation radiation monitors is 24 vdc.
. 0 h.. -.
Request that consideration be given during grading to answers indicating - that IRH "A" and the HSIVs will be affected by a manual transfer of RPS i Bus "A".
During the loss of power to RPS Bus "A", the RPS channel "A" relays which are deenergizad by IRH "A" scram signals would be deenergized.
An IRH High High/IN0P alarm would annunciate and a half scram would occur.
The operator would have to reset the half scram when power returned. Also, during the loss of power, PCIS logic channel "A" is deenergized, resulting in a half isolation signal to the Group I isolation logic.
This would require the operator to reset the logic aftt.r power returns.
REFERENCE: PNPS Reference Text - 480 V System, pp. 13-1 and 14-1 PNPS Reference Text - PRM System, p. 37-1 PNPS Reference Text - PRM System, pp. 33 and 34 .
.- 18780/21 QUESTION 2.08: (3.00) - The RCIC System started on an automatic initiation signal and has been operating for 5 minutes.
For each of the followir.g conditions, STATE whether the RCIC system HILL or HILL NOT continue to operate.
If it will continue to operate, HILL there be any adverse effects from RCIC operation under the conditions? If it HILL NOT continue to operate, HHY NOT? Consider each condition separately. Assume no operator action.
a. The condensate pump for the barometric condenser fails causing a high level in the barometric condenser.
(1.00) b. Reactor pressure decreases to 100 psig.
(1.00) c. The RCIC lube oil pump fails causing oil pressure to drop to 1 psig.
(1.00) 6NSHER:
a. Hill continue to operate.
(0.50) Operation under these condi-tions would allow contamination of the RCIC room and atmosphere from turbine and valve steam leakage.
(0.50) b. Hill continue to operate.
(0.50) No adverse effects on RCIC operation.
(0.50) (Isolation is at 67 psig.)
c. RCIC will not continue to operate.
(0.50) RCIC turbine will trip due to insufficient oil pressure to hold trip and throttle valve open.
(0.50) REFERENCE: Pilgrim: Systems Reference Texts, RCIC, pp. 3, 4, 5 and 19, Figure 6.
KA 217000K405 (3.2/3.5) 217000A207 (3.1/3.1) 217000A104 (3.6/3.6) 217000G07 (3.8/3.7) COMMENTS: l
a. NO COMMENT , b. Request that the answer be changed to "RCIC will not continue to operate.
(0.50) RCIC will isolate on low reactor pressure (any isolations also trips the turbine).
(0.50)" PHPS Technical Specifications, Table 3.2.B states that the RCIC steam line low pressure isolation setpoint is 100 > P > 50 psig.
Also, the RCIC student guide state that the low pressure isolation is set between 50 and 100 psig.
This is what the trainees were taught.
18780/22 c. Request that two answers be considered.
- 1. If the trainees assumed that the control valves opened far enough on loss of oil pressure to overspeed the turbine, request that an answer stating that the turbine will not continue to operate because of an overspeed trip be accepted.
2. If the trainees assumed that the turbine control valves opening during RCIC injection does not cause the trubine to trip on overspeed, request an answer stating that the turbine will continue to operate with probable bearing damage be accepted.
PDCR 81-48, RCIC Automatic Restart Modification, modified the turbine trip and throttle valve to prevent it from tripping on a loss of oil pressure.
(The dashpot trip mechanism was replaced by an electrical trip solenoid).
The turbine trip and throttle valve is no longer affected by a loss of lube oil pressure.
This is what the trainees were taught.
REFERENCE: Pre-startup Training Manual, March, 1982, Part II, PDCR 81-48, page 42/58.
Reactor Core Isolation Cooling Student Guide, page LG-il., PNPS Technical Specifications, Table 3.2.B >
_ _ _ _ _ __.
__ _ . _ _ _ _ _ - _ _ _ _ - -. 18780/24 QUESTION 2.10: (2.50)
Concerning the Standby Gas Treatment System (SBGTS): a. When an initiation signal occurs, the dampers to the exhaust plenum open. Air is drawn from several locations. Refer to the attached SBGTS Figure 1 and LIST the three (3) locations from which air can be drawn which corresponds to the blanks on the Figure labelled 1, 2 and 3.
(1.50) b. If the SBGTS was automatically initiated by a reactor low water level signal, WHAT must be done to shut the system down once the low-level signal clears? (1.00) ANSWER: a. 1. Refuel floor (0.50) 2. Drywell (0.50) 3. Suppression pool (0.50) b. The drywell isolation reset button must first be reset.
(1.00) REFERENCE: Pilgrim: Procedure 2.2.50, p. 10.
Pilgrim: Procedure 2.4.147, p. 3.
Pilgrim: Systems Reference Texts, SGTS, pp. 3 and 12 and Figure 1.
KA 261000K101 (3.4/3.6) 261000K102 (3.2/3.4> 261000K103 (2.9/3.1) 261000K401 (3.7/3.8) COMMENTS: a. Request that the sequence of points 1, 2 and 3 be allowed to be given in any order. Confusion could have existed in the candidates attempting to correctly identify points 1, 2 and 3 because the supplied drawing is in error.
It was noted that A0-N101 is actually in the refuel floor suction line rather than in the drywell vent line as indicated.
The error has been previously overlooked because the supplied drawing is generally used only to support discussion of the SBGT train components.
The interrelationships between SBGT suction paths and reactor building ventilation are discussed in detail during the Reactor Building Ventilation System lecture.
The confusion would be a result of candidates knowing that the refuel floor has no direct conr.ection to the SBGT plenum. Additionally, they understand that both the drywell and torus have separate direct connections to the plenum with no means for being cross connecting to clean exhaust.
Also, we request that Primary Containment Atmosphere Control (PCAC) be considered as an alternate answer to drywell and suppression pool.
The student may include drywell and suppression pool by saying PCAC since the drywell and suppression pool vent and purge lines are commonly referred to as part of the Primary Containment Atmosphere Control System.
- - . -
. ___ ______.
18780/25 b. Securing Standby Gas Treatment after automatic initiation, requires - resetting of the isolation signals at panel 905 and panel C-7 in accordance with PNPS Procedure 2.4.147, Reset of Secondary Containment Isolation on Panol C-7 Please consider, as an alternate answer, statements to the effect of resetting the isolation and restoration of the Standby Gas Treatment System lineup.
REFERENCE: Reactor Building Ventilation Reference Text Figure I and 12 HVAC drawing H-294 Primary Containment Atmosphere Control System Student Guide, 0-R0-02-08-02, page LG-2 Primary Containment Atmosphere Control Reference Text, Figure 5 Standby Gas Treatment System Student Guide, 0-RO-02-08-03, page LG-8 PNPS Procedure 2.4.147, pages 3 and 4.
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. .~. . 18/o0/26
OUESTION 3.01: (3.00) Following a reactor SCRAM, some scram signals are bypassed by oper-ator or automatic actions.
For each of the following scram signals, STATE all the condition (s) that must be in effect for a bypass to occur: (3.00) a. Main steam line isolation scram b. Reactor mode switch in SHUTDOWN scram c. Turbine control valve fast-closure scram d. Scram discharge volume high level scram ANSHER: a. Bypassed when the mode switch is NOT in RUN.
(0.75) b. Auto bypcased after (2 sec.) time delay.
(0.75) c. Auto bypassed if reactor power < 45 percent (as indicated by turbine first stage pressure of 305 psig).
(0.75) d. Manual bypass switches in BYPASS with mode switch in SHUTDOHN or REFUEL.
(0.75) REFERENCE: Pilgrim: Systems Reference Texts, RPS, pp. 17 through 25.
KA 212000K412 (3.9/4.1) 212004K408 (4.2/4.2) COMMENTS: a. Request that the answer be changed to include reactor pressure being < 600 psig as part of the answer.
To bypass the main steam line isolation scram, the mode switch must be "NOT IN RUN" and reactor pressure must be < 600 psig.
Failure to meet either of these criteria will make the main steam fine isolation scram available, c. Request that either part of the answer, less than 45% or 305 psig ist stage pressure be accepted as an answer.
The instrument that provides the bypass signal for the turbine control valve fast closure scram is a turbine first stage pressure detector.
REFERENCE: Reactor Protection System, Student Guide, 0-R0-02-07-02, Figure 8 .
18780/29 QUflUDN 3.04: (2.00) - Consider the reactor water level instrumentation: a. For each of the following parameter changes and operational con-ditions, STATE whether the INDICATED vessel level will INCREASE, DECREASE or REMAIN THE SAHE for the specified level instrument.
ACTUAL vessel level REMAINS THE SAME.
1. The reactor vessel temperature increases from 120 degrees F to 200 degrees F during a reactor startup.
H0H will the SHUTDOHN RANGE level instrumentation respond? (0.50) 2. The reactor is in cold shutdown.
The reactor recirculation pumps trip.
H0H will the FUEL ZONE level ir.strumentation respond? (0.50) b. WHICH water level instrument RANGE provides the low water level trip inputs to the RPS logic? (0.50) c. WHICH water level instrument RANGE provides the water level inputs to RHR Containment Spray logic? (0.50) , ANSWER: a.1. Increase (0.50) 2. Decrease (0.50) b. Narrow range (0.50) c. Fuel zone (0.50) REFERENCE: Pilgrim: Systems Student Guide: Non-Nuclear Instrumentation, pp. 23 through 25.
Pilgrim: Systems Reference Texts, Nuclear Boiler Instrumentation, pp. 5, 6 and 22.
KA 216000K501 (3.1/3.2) 216000K507 (3.6/3.6) 216000K510 (3.1/3.3) 291002K108 (2.8/2.9) 216000K101 (3.9/4.1) / COMMENTS: a. 1. Request change answer to decrease.
130 degrees F to 200 degrees F increase implies the reactor is heating up.
Page 27 of the Student Guide states the. wide range level instruments will read lower than actual as thu reactor heats up.
This is supported by the discussion concerning differential pressure level instruments in the Instrumentation and Control academic material.
18780/30 a. 2. Request alternate answer of "no change".
- During this condition, the level will be greater than +22.5'
inches, therefore, you will not be able to see any level change, instrument will be pegged high.
b. Request alternate answer of - 50 to + 50 This is the actual RANGE of the narrow range instrument.
c. Request alternate answer of - 50 to + 50 (narrow range instrument).
The LPCI initiation logic requires 1) 1 -49 inch signal (Narrow Range Instrument); and 2) 1 400#. A LPCI initiation signal will close the containment spray valves.
The < 2/3 core coverage input to the logic comes off the fuel zone instrument.
REFERENCE: PNPS Non-Nuclear Instrumentation, Student Guide, pp. 23-24, 27 PNPS Instrumentation and Control, Student Text, pp. 27-33 l l ! i ! J'
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18780iSi ' OUESTION 3.05: (2.00) The reactor is operating at 70 percent power.
Flow converter A fails such that its output is downscale.
STATE ALL the CAUSE(s) for each trip which will occur.
(2.00) ANSHER: R00 BLOCK - due to flow comparator INOP (0.50) - due to flow biased trip (70 > [0.58H] + 50; H-0) (0.50) - RBM flow biased trip (0.50) ONE-HALF SCRAM Ch. A - due to flow biased trip (70 > (0.58H] + 62; H-0) (0.50) REFERENCE: Pilgrim: Systems Reference Texts, APRM, pp. 4, 5, 8, 9 and 19.
KA 215005Kil6 (3.3/3.4) 215005K607 (3.2/3.3) 215002K101 (2.9/3.0) 212000K603 (3.5/3.7) COMMENTS: For the rod blocks which are caused as a result of this malfunction, request that a "Flow Converter Hismatch" be added as an additional acceptable response.
REFERENCE: i PNPS APRM Reference Text, p. # APRM 4-2 l l J' l --
18780/32 , ' ' OUESTION 3.06: (3.00) Consider the Rod Block Monitor System (RBH): a. WHAT is the purpose of the null sequence control circuit? (1.50) b. WHAT are three (3) ways the RBH trips can be BYPASSED? (1.50) ANSHER: a. Adjusts the gain of the RBH channel (0.75) to compensate for variations in local power (0.75) OR bypassed LPRH's.
(0.75) b. 1. Joystick (0.50) 2. Edge rod selected (0.50) 3. Reference APRM < 30 percent (0.50) REFERENCE: Pilgrim: Systems Reference Texts, RBM, pp. 3, 6, 8, 15 and 18.
KA 215005K502 (2.4/2.5) 215005K403 (2.9/3.0) 215005A403 (2.8/2.8) COMENTS: a. Request that points be reassigned to allow full credit for either alternative answer to part two.
The answer to part "a" appears to be divided up into two parts with the last part having two alternative answers.
However, the point assignment requires both alternatives for the second part to receive full credit.
b. Request that reactor power < 30% be accepted as an alternhtive answer to "reference APRH < 30%". PNPS Technical Specifications state that the Rod Block Monitor rod l blocks are bypassed when reactor power is 130% power. r
REFERENCE: RO exam answer key PNPS Technical Specifications, Table 3.2.C, note 7
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18780/33 ' 00ESTION 3.07: (3.00) Consider the Process Radiation Monitoring (PRM) System: a. Specifically, HHERE, in relation to the HSIVs, are the main steam line (HSL) radiation monitoring detectors located? (1.00) b. On a trip of the HSL radiation monitors, HHAT three (3) automatic actions, besides the closure of the HSIV's, could occur as a direct result of the HSL Hi Hi Rad signal? (1.50) c. HHAT, if any, automatic action (s) occurs if one channel of off-gas PRM fails high while the other channel of post-treatment off-gas PRM is downscale? (0.50) ANSHER: a. Detectors are located in the steam pipe tunnel, next to main steam piping, downstream of the outboard MSIV's.
(1.00) b. 1. Reactor scram (0.50) 2. Mechanical vacuum pump isolation (trip) (0.50) 3. Main steam line drains close (0.50) c. Trip is initiated (13-second timer starts) and off-gas hold-up volume isolates.
(0.50) REFERENCE: Pilgrim: Systems Reference Texts, PRM, pp. 6 and 22.
KA 272000K101 (3.6/3.8) 272000K102 (3.2/3.5) 272000K402 (3.7/4.1) COMMENTS: a. Request that "downstream of the outboard MSIV's" be placed in parentheses.
b. Request that "recirculation sample valves" be included as one of the possible answers.
c. Request rewording of answer to "off-gas holdup line will isolate after the 13 minute timer times out."
(It is a 13-minute timer) REEEBLN.CE: PNPS Primary Containment Student Guide, page LG-19 PNPS Procedure Radiation Monitoring Student Guide, page LG-5 (See attached reference) l
_ ___ 18780/34 OUESTION 3.08: (2.00) ' The reactor is operating at 100 percent power with the following Mechanical Hydraulic Control (MHC) System setpoints: EPR setpoint: 930 psig HPR setpoint: 933 psig Bypass Opening Jack setpoint: O percent (BPV's are closed) Speed / Load Changer demand: 100 percent (rated load position) Briefly DESCRIBE the HHC System response to the following twc separate situations.
INCLUDE in your discussion any changes in reactor pressure, control valve position and bypass valve position and HHY these changes occur. Assume no operator action.
a. The EPR fails so that reactor pressure decreases.
(1.00) b. Power is lost to the EPR.
(1.00) ANSHER: a. Reactor pressure is decreasing (given) due to EPR causing the control valves to open.
(0.33) BPV's remain closed.
(0.33) Reactor pressure continues to decrease until HSIV's isolate on low steam line pressure (with the mode switch in RUN).
(0.34) b. On power loss of EPR, EPR closes the control valves which causes reactor pressure to increase.
(0.33) When reactor pressure reaches the HPR setpoint, the HPR will control the control valves to maintain pressure.
(0.34) BPV's do not open.
(0.33) REFERENCE: Pilgrim: Systems Student Guide, Main Turbine Steam System, pp. 16 through 19.
KA 241000K419 (3.6/3.7) 24100A102 (4.1/3.9) 24100A107 (3.8/3.7) 24100A308 (3.8/3.8) COMMENTS: e a. Request that the answer be changed to state that the bypass valves will open.
If the EPR fails such that pressure is decreased, the control valves and bypass valves will open and depressurize the reactor.
Control valve position will be limited to 105% rated steam demand. Hith the maximum combined flow limiter set at 110% rated steam demand, the bypass valves will open approximately 5%. b. NO COMMENT REFERENCE: Main Turbine Steam System Student Guide, 0-R0-02-05-01, pages LG-21 and LG-39.
18780/37 ' 00ESTION 4.01: (2.00) The reactor is operating at full power and the following alarms occur: PANEL C-7 TROUBLE ALARM 'and ORYHELL AIR COOLER HIGH ORAIN FLOH ALARH In accordance with Procedure 2.4.14, Leaks Inside Primary Containment.
List four (4) possible symptoms (other than alarms) that you would observe prior to a reactor scram.
(2.00) ANSHER: 1. Excessive sump pump operation due to increased leakage to drywell (0.50 equipment drain sump or drywell floor drain sump.
each) 2. Abrupt increase in drywell humidity (Panel C85).
3. Significant drywell pressure increase.
- 4. Significant drywell temperature increase.
REFERENCE: Pilgrim: Procedure 2.4.14, Leaks Inside the Primary Containment, p. 2.
KA 295010AK103 (3.2/3.4) 295010AA106 (3.3/3.5) 295010AA202 (3.8/3.9) 295010AA204 (2.8/3.0) COMMENTS: Consider as an additional answer, RBCCH low discharge pressure.
Common operating practice would be to check RBCCH even though it is not listed in Procedure 2.4.14.
Using all possible indications has been stressed to the trainees, r
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' OUESTION 4.02: (3.00) a. According to Procedure 2.4.23. "Jet Pump Flow Failure", unexplained changes in jet pump flow could indicate jet pump failure. HHAT are two (2) other indications or symptoms of jet pump failure? (1.50) b. HHAT two (2) actions would you take upon observing the abo'.'e symptoms? (1.50) ANSHER: a. 1. Recirculation system delta pressure deviation (as determined by test).
2. Sudden change in core delta pressure without accompanying change in recirculation pump speed.
3. Unexplained changes in recirculation flow.
Any two (2) (0.75) each, (1.50) maximum, b. 1. Monitor alarms and instrumentation to determine type of system malfunction that has occurred.
(0.75) 2. If it appears that a jet pump problem exists, refer to the Technical Specification LCO.
(0.75) REFERENCE: Pilgrim: Procedure 2.4.23, p. 2 KA 202001K601 (3.5/3.7) COMMENTS: a. Request the core plate AP be excepted as an alternate answer for core AP rynonymous terminology.
b. NO COMMENT r REFERENCE: PNPS Reference Text Nuclear Boiler Instrumentation (page 3-3).
18780/46 OUESTION 4.10: (1.50) - PROVIDE the following action levels or limits for plant operation: a. Maximum allowable temperature difference between the vessel dome and bottom head drain just prior to starting a recirculation pump.
(0.25) b. Maximum allowable temperature difference between recirculation loops just prior to starting a recirculation pump.
(0.25) c. Maximum permitted cooldown rate during a normal shutdown.
(0.25) d. Maximum pressure differential across an MSIV prior to opening the MSIV.
(0.25) e. Maximum permitted RPV shell flange to shell temperature differ-ential.
(0.25) f. Maximum power level for mechanical vacuum pump operation.
(0.25) ANSWER: a. 145'F (0.25 b. 50*F each) , c. 100*F/hr d. 200 psig e 145'F f. 5 percent REFERENCE: Pilgrim: Procedure 2.1.9, p. A-2.
Pilgrim: Technical Specifications, Section 3.6.A.
Pilgrim: Procedure 2.2.92, Precaution A, p. 7.
. Pilgrim: Procedure 2.2.93, Precaution 0, p. 7.
l Pilgrim: Procedure 2.1.3, Limitations A.3, B.2, pp. 2 and 3.
COMMENTS: ) a. NO COMMENT b. NO COMMENT c. NO COMMENT d. Request that consideration be given to accepting 50 psid as the actual differential pressure limit across MSIVs prior to opening of the valves.
Precautions in PNPS Procedure 2.2.92, caution the operator to not l attempt to open an MSIV with a differential pressure of > 200 psid j across it.
This is a design parameter and prevents valve damage.
Step 7 of the procedure for opening MSIVs with the reactor pressurized states a conservative operational differential pressure limit of 50 psid across the MSIVs.
. - - .
18780/47 ' e. NO COMMENT f. NO COMMENT REFERENCE: PNPS Procedure 2.2.92, Step A7 x J'
J
_ - _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _-. _ . - _ _ _ _ _ _ _ _ _ _ _ _ . ATTACHMENT 4 NRC Resolutions of Comments on. Written Examinations.
The following represents the facility comments and the NRC resolution to those comments made as a result of the current examination review policy.
Only those comments resulting in significant changes to the master key, or those that were not accepted by the NRC, are listed and explained below.
Comments that were insignificant in nature and resolved to the satisfaction of both the examiner and the licensee during the post examination review are not listed.
i.e.: typo errors, relative acceptable terms, minor set point changes SR0 EXAMINATION Question 5.04: Comment accepted Question 5.06: Comment accepted.
Parenthesis placed before "and control..." to the end of the paragraph.
Question 5.09: The candidate is expected to use information obtained in the previous section. This knowledge will be graded proportionatly (.1 pts.) with methodology on how to calculate the moderator temperature change (.9 pts.)
Comment partially accepted Question 6.01: a.
According to the reference material provided on PNPS RCIC Study Guide p. LG-11, the 300% steam flow signal must be provided for greater than 3 seconds.
Comment not incorporated b.
Comment incorporated (typographical) l l l ! l l l l .
. _ _ , . Attachment 4
Question 6.05: a.
Comment accepted.
Parenthesis to be located before "after a design...". Question 6.08: a.
Comment accepted. Additional reference material reviewed and accepted.
Question.6.09 Comment accepted. Additional reference material was reviewed and accepted answer.
Flow converter Inop Trip will be incorporated as an optional answer for 10% flow mismatch.
Full Credit will be awarded (0.5) for either response.
Reference PNPS APRM Ref. Text p. 4-2.
Question 7.01 a.
NRC review of the question has determined the correct answer to be 107 inches. A tolerance of "plus or minus" 3 inches will be added to the answer key to account for interpolation, b.
Comment accepted. A band of 50 psig'"plus or minus" 25 psig will be acceptable.
c.
Comment partially v:cepted. Candidate will be graded on the basis of WHY the actions stated in E0P-3 are taken, as referenced in E0P's Cautions, Variables and Curves enabling learning objective 10 for the Heat Capacity Level Limit and by K.A 295030 EK 1.01 and 1.03.
Full credit (1.0) will be given for a response similiar to that in the last sentence of answer 7.01.c.
, ! l-Question 7.02 Comment accepted.
Typographical error caused the deletion of temperature.
It will be added as option e with only 4 of 5 responses required for full credit.
Question 7.04 c.
Comment accepted. Added "(to bottom of separator skirt)" to the j answer for clarity.
Reference page PNPS 2.4.25-3.
L l-Question 7.07 b.. Comment accepted.
For full credit candidate must discuss increase in blowdown area as part of the answer.
Points will be redistributed (0.5) for blowdown area, (0.5) for 2/3 core coverage.
Reference Bases of T.S. 3.6.E and 4.6.E.
E a.
. - -.. -. - , . - -, _ - - - _. - - - - -.. - .-, - .
- ... _ . . - - - . Attachment 4 '3 Question 7.11 b.
Comment accepted. Correct response change to 50'psid upon review of additional reference provided.
Reference PNPS 2.2.92.
Question 8.01 Comment not accepted.
Sect. C and D cannot be performed immediately and therefore do not answer the stated question.
Reference T.S. 6.7 Question 8.08 a.
Comment not incorporated.
Existing answer key provided for that option.
Question 8.09 a.
The problem of collet housing failure is one example of a generic control rod drive problem.
Full credit will not be given on that single response, though a discussion of the problem is relevant to.
the answer and will be considered for partial credit.
Comment not incorporated.
Reference T.S. Bases 3.3 and 3.4 p. 88.
R0 Examination Question 1.03 b.
Comment accepted. Alternate answer b.1. is acceptable only if the candidate explains why the ratio of local flux to average flux decreases when void content increases.
Question 1.04 Comment accepted.
Question 1.09 Comment partially accepted. The second requested additional alternate answer "Procedural requirement for water level above 35 inches prior to
pump start" is NOT to ensure ade: ' ate NPSH and therefore is not considered to be an alternate answer.
Question 1.11 Comment accepted.
Question 2.01 a.
Comment accepted.
.
. ._ - - - .. , . . Attachment 4
Question 2.01 b.
Comment not accepted. The Student Guide for RBM is 0-RO-02-07-01, "NMS."
Therefore learning objectives #6, 7, and 13 apply to the RBM.
Also Procedure 2.4.38, "LPRM Failure," step IV.F. states, "If the failed LPRM is feeding the RBM, verify that the. sufficient inputs.to the-rod blocks are maintained.
The RBM will become inoperable if less than 50% of its inputs are available." The step does not tell the operator.how to verify the number of inputs. The operator is expected to know how to perform this step.
If necessary, the learning objectives should be clarified or expanded to include the above knowledge.
Question 2.02 c.
Comment accepted.
The "or" in the answer key should have been capitalized to identify either action is an acceptable response.
Question 2.04 a.
Comment partially accepted. The alternate answer not accepted because the alarm does not occur at high or high-high level, as required by the-question. The high level annunciator in the control room was given in the question.
Question 2.06 Comment accepted. The comment identifies synonymous terminology and hence no change to the answer key is required.
Question 2.07 a.
Comment accepted.
i Question 2.07 b.
Comment partially accepted. Answer b.3 (Reactor building ventilation radiation monitors) is deleted and replaced by "IRM A."
The MSIVs (components) are not affected and therefore not an alternate answer.
System Reference Text, Reactor Protection System (including Figure 1) should be revised to reflect the fact that the reactor building ventilation radiation monitors are no longer powered from the RPS buses.
Question 2.08 b.
Comment accepted. The Systems Reference Text (RCIC) should be changed to identify the flexibility allowed by Technical Specifications Table 3.2.B.
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_ ._ _ _ Attachment 4
Question 2.08 c.
Comment partially accepted. _The acceptable answer is that the control valves will open and overspeed the turbine (overspeed trip) and hence the turbine will not continue to operate. No information was provided to-substantiate any assumption that the overspeed would not cause an overspeed' trip.
t Question 2.10 a.
Comment partially accepted. The locations from which the air is drawn may be listed in any order.
The Systems Reference Texts (SBGTS) Figure 1 should be corrected.
The Primary Containment Atmosphere Control System is not an acceptable alternate answer because the question asked for "location" not a "system."
Question 2.10b.
Comment accepted. The acceptable answer is that the drywell isolation < ' must first be reset.
Question 3.01 a.
Comment accepted. The acceptable answer includes "and reactor pressure less than 600 psig."
Question 3.01 c.
Comment accepted.
Either expression of the setpoint identified in the answer key is acceptable. No change to the answer key is required.
Question 3.04 a.
Comment accepted.
Question 3.04 b.
i Comment accepted.
! Question 3.04 c.
Comment partially accepted.
The acceptable answer requires both "Fuel Zone" and "Narrow Range (-50 to +50 inches)."
Question 3.05 . Comment partially accepted.
"Flow converter mismatch" is an alternate answer to "flow comparator INOP." Both give the "Flow Reference Offnormal" annunciator on Panel 905.
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. Attachment.4 6_ . . Question'3.06 a.
~
- Comment accepted. The point assignment for each part of the-answer has been clarified.
' . Question 3.06 b.
Comment accepted. No change to the answer key is required.
Question 3.07-a., b., c.
Comments accepted.
Question 3.08 a.
Comment accepted.
Question 4.01 Comment not accepted.
The stated annunciator (Drywell Air Cooler High Drain Flow Alarm) is not indicative of an RBCCW 1eak. The question asks what "would" be observed, not what "could" be observed. Also, the question specifically states, "In accordance with Procedure 2.4.14."
Question 4.02 a.
I Comment accepted. The comment identifies synonymous terminology and hence no change to the answer key is required.
Question 4.10 d.
comment accepted. The acceptable answer is 50 psid based on the PNPS procedural limit.
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