ML20150E098
| ML20150E098 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 03/21/1988 |
| From: | Mccrory S, Pellet J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20150E095 | List: |
| References | |
| 50-285-OL-88-01, 50-285-OL-88-1, NUDOCS 8803290324 | |
| Download: ML20150E098 (78) | |
Text
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APPENDIX U.S. NUCLEAR REGULATORY COMMISSION REGION IV l
l l Operator' Licensing Exam Report: 50-285/0L-88-01 Operating License: DPR-40
- 1. l l
Docket No: 50-285 Licensee: Omaha Public Power District 1623 Harney Omaha, Ne 68012 i Facility Name: FortCalhounStation(FCS)
Examination at: Fort Calhoun Station Chief Examiner: -// . M/
- 5. t'.' McCrory, lead Examipr, Date /
Operator Licensing Sectio'n, Division of Reactor Safety Approved by: & L 3 ll; N J.*L. Pellet, Chief, Operator Date ,
Licensing Section, Division !
I of Reactor Safety l
Sumary NRC Administered Examinations Conducted During the Week of February 8, 1988 (Report 50-285/0L-88-01)
Results: NRC administered examinations to 3 candidates. Three (3) candidates passed all portions of the examination and have been issued the appropriate license.
8803290324 880322 PDR ADOCK 05000285 V DCD _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
DETAILS
- 1. Persons Examined SR0 R0 Total .
License Examinations: Pass - 1 2 3 Fail - 0 0 0
- 2. Examiners S. L. McCrory, Chief Examiner L. Defferding, PNL J. L. Pellet, NRC
- 3. Examination Report Performance results for individual examinees are not 'ncluded in this report as it will be placed in the NRC Public Document Room and these results are not subject to public disclosure.
- a. Examination Review Comment / Resolution In general, editorial comments or changes made during the examination, or subsequent grading reviews are not addressed by this resolution section. This section reflects resolution of substantive comments made by FCS. The only comments addressed in this section are those which were not accepted for incorporation into the examination and/or answer key. Those comments accepted are incorporated into the master examination key which is included in this report. -Comments may be paraphrased for brevity. The full text ,
of the coni,ents is attached.
(1.05) "Increase temperature" should be considered an ;
acceptable answer.
Response: Reject. The operator cannot simply increase temperature directly. There are several actions which result in a temperature increase for which credit can :
be given. However, to allow "increase temperature" to be acceptable is to disassociate results from actions.
(1,10/5.10) The first sentence alone is sufficient to answer the question.
Response: Reject. This is FCS Question 007-007-01-C-16-1.02-0002 The answer is structured to indicate that the facility licensee expects the entire response shown to properly answer the question.
, . - - - _ _ _ . - _ _ . ,, .r_.. - . . __- .e.__ c_ - , , _ , _ , ,....___....-mr--_,,_,---.-._.-.-...w_, ,_, - c
(5.03) Two additional answers should be added to the key.
Response: Reject. This is FCS Question 007-005-05-C-03-3.04-0006.
The proposed additional answers contradict the original key.
(8.04) "Completion of the work" should be considered sufficient to get full credit.
Response: Reject. Per the 5.0 , "Flame ... Permits," are governed by M0s and DCOs. Therefore "completion of work" is inadequate. This is FCS Question 007-015-08-C-05-0.07-0001.
- b. Site Visit Sumary (1) At the end of the written examination administration, the facility licensee was provided a copy of the examination and answer key for the purpose of commenting on the examination content validity. It was explained to the facility licensee that regional policy was to have examination results finalized within 30 days. Thus, a timely response was desired to attain this goal.
(2) At the conclusion of the site visit, the facility representative determined that an exit meeting was unnecessary since no results could be reported until finalized.
- c. Generic Coments Several questions in the examinations were taken from the FCS exam question bank furnished to the NRC. In addition to those noted in Section 3.a above, several more were identified in the facility licensee review of the NRC prepared examination as having technical errors in the answer key. The number of these occurences indicates a lack of adequate technical review of facility licensee generated questions for accuracy. NRC suggests that the facility licensee conduct a comprehensive review of the FCS (licensed operator) question bank to correct technical inaccuracies and provide a revised question bank to Region IV Operator Licensing Sectior.
- d. Master Examination and Answer Key Master copies of the FCS license examinations and answer keys are attached. The facility licensee coments which have been accepted are incorporated into the answer key.
- e. Faciltiy Examination Review Coments The facility licensee comments regarding the written examination are attached. Those coments which were not acceptable for incorporation into the examination answer key have been addressed in the resolution section of this report.
(
U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _EI._C6LBOUU- = = _ _
REACTOR TYPE: _ EWE =CE ___
DATE ~ ADMINI STERED : _SSZG2Za2----
EXAMINER: _BCCEDEY._S.____--
CANDIDATE: _________ __________
luSIEUCIlCUS_IQ_CAUDineIE.:.
Use separate paper for the answers. Write answers on one 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.
% OF CATEGORY % OF CANDIDATE'S CATEGORY
__VeLUE_ _IDIAL ___ SCORE ___ _ U e L U E _ _ _ --- -=_ feIEGDEX_- - - -
_25.00__ _25.00 _ _ - - - = - - __ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS
_25.Qa__ _25.Ca _ _ _ _ _ - - -
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION
_25.aa__ _25.QQ ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
_25.Qa__ _25.Qa ___________ ________ S. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS ICQ.QQ__ ___________ __ __% Total s Final Grade All work done on this examination i s my own. I have neither given nor received aid.
Candidate's Signature i
L
5m_$IBEDEY_DE_UUCLE6E_EDWEE_ELsUI_DEEEellOUo_ELUIDS _eUD PAGE 2 IBEEMDDYU6 MIDS QUESTION 5.01 (1.50)
Whfle operating cri tical below the point of adding heat at-EOL, the reactor operator pulls control rods to establish a constant 0.5 DPM startup rate.
How much reactivity was added? (Show all work and state assumptions.)
ANSWER 5.01 '1.50)
SUR = 26.06/T T = 52.12 SEC (0,5) rho = (Beta-bar-effective 3/(1+ lambda T) = 0.0055/1+.1x52.12 (0.5)
= 0.00088 d tno = 0.088% drho (.080 .095%) (0.5)
REFERENCE FCS RXTH LP 7-5-4 EO 4.3 192003K106 ...(KA'S) l QUESTION 5.02 (2.50)
- a. What is (are) the source (s) of decay heat follcWing a reactor i shutdown? (0.5) l l b. What factor (s) affect the magnitude of decay heat? (1.0)
I
- c. Why is RCS heat removal capability important with respect to decay
- heat? (0.5) l d. Why is the power level represented by decay heat not indicated by the nuclear instruments? (0.5) l l
l l
l 5 __IBEDEX_DE_UUCLEAE_EDWER_ELANI_DEEEAIlOU4 _ELUIDS4_aMD PAGE 3 IBEEUDDYUaUIDS ANSWER 5.02 (2.50)
- a. Decay heat is produced from the decay of fission products (fragments) formed during power operation. (0.5)
- 6. Power history (0.5) and time since shutdown (0.5)
.c. To prevent fuel clad damage (failure) (0.5)
- d. The bulk of the decay heat is produced from fission products which do not emit neutrons (alpha, beta, or gamma emitters) as part of the decay process. (0,5)
REFERENCE FCS RXTH 007-005-04 192008K127 ...(KA'S>
QUESTION 5.03 (1.50)
If reactor power is kept constant wnile control rods are withdrawn and boron concentration is increased, cxplain the effect on linear heat rate (kW/ft) during such a maneuver.
ANSWER 5.03 (1.50)
Linerar heat rate will decrease. (0.5)
As the rods are withdrawn, more reactor fuel is available to generate the same power thereby decreasing the power density. (1.0)
REFERENCE FCS RXTH LP 7-5-3 EO 3.4 001000K501 ...(KA'S)
(+++o C AT E G O R~( 05 CCNTINUED ON NEXT PAGE +++++)
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5._iIBEQEY_DE_NUCLEGR_EQUEE_ELeUI_0BEEellON# _ELulDS._aSQ PAGE 4'
,' 'IBEEMODYMAMICS 4
OUESTION 5.04 (1.50)
Assume the follo41ng. conditions:
BOC with a' sl i gh t l y positive MTC Reactor power 16% ;
Steam dumps in manual i.
Turbine control in manual !
I !
- - A peripheral rod drops during these conditions. Briefly describe ~ and i explain _the reactor response to this event assuming NO operator action. l t ;
i I
1 I !
! ANSWER 5.04 (1.50) :
- . The dropped rod will cause reactor power to decrease (go subcritical) due
! to the negative reactivity insertion. (0.5) Since steam demand will remain l f constant, the RCS will begin to cooldown. (0.5) With a positive MTC, more ~
L negative reactivity will be added to the core as RCS temperature continues [
to drop. (0.5) (Ultimately RCS temperature will stabilize' at the heat !
addition capacity of the RCFs combined with a reduced steam rate based on l ower secondar y temperature and pressure.)
s REFERENCE I FCS RXTH 007-005-03 192005K103 ...(KA'S) ;
i QUESTION 5.05 (1.50) l The reactor is critical below the point of adding heat with rod control in manual ont EOL conditions when a steam dump f ail s open. Expl ai n how e ach of f the f ol l on i ng p ar ame t e r s is affected. Continue the answer until stable i conditions are reached. (Assume no operator actions and no protective [
function actuations.) Values are not required. j
- a. Tavg ;
- b. Fuel temperature !
L c. Reactor p owe r f I '
I L
l h
i I !
l !
' ' 5.
- TBEDRY_DE_tlUDLEAE_EDWFP ELMT .DEFRATlDS_ELijlDS,_etJD PAGE 5 IBERUDDXNAHTDS ANSWER 5.05 (1.50)
- a. Tavg decreases on an excess steam fl ow un t il reactor power increases to match: steam flow
- b. Fuel temperature will increase to the point of adding heat and turn power.
- c. Power will-' increase to match steam flow.
(0.5 ea)
REFERENCE FCS RXTH LP 7-5-4 EO 5.5 001000K517 ...(KA'S)
OUESTION 5.06 (1.50) ,
Why do technical specifications impose both an upper and lower limit to the magnitude of MTC?
ANSWER 5.06 (1.50) [
The upper limit (positive MTC) is set to ensure reactor stability. (0.5)
The lower limit (large negative MTC) reduces tne impact of a postulated j
steamline break with subsequent cool down and poten t i al for return to power. l (CONCEPT) (1.0)
REFERENCE FCS RXTH 007-005-03 192004K106 ...(KA'S)
QUESTION 5.07 (1.50) i Oln a reactor trip from equilibrium Xenon conditions, how do the differential Xenon reactivity insertion (peak worth - equilibrium worth) and time to peak insertion vary as equilibrium conditions are changed from L 100% to 20% power?
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, ' L 'TBEDEY_DE_UUCLFAR EDUER_ELeNI_DEERellDtd_ELUIDS4 _6ND PAGE 6
-IBERdDDYNAMicS ANSWER 5.07 (1.50)
The ' di f f eren t i al reactivity insertion decreases (0.75) as does.the time to peak insertion (0.75).
REFERENCE FCS RXTH 007-005-03 000005K103 001000K538 ...(KA'S) i QUESTION 5.08 (1.00)
What is the basi s f or_ al l owi ng a reduction in the minimum shutdown margin requirement once RCS temperature goes bel ow 210 deg. F?
ANSWER 5.08 (1.00)
'The potential for a reactor restart due to a steam line rupture and cooldown no longer ex i s t s be l ow 210 -de g. F. (1.0)
REFERENCE I
FCS'RXTH 007-005-06 192002K114 ...(KA'S) i i
! QUESTION 5.09 (1.00)
The power operated relief valve (PORV) opens whlie at power. The
! pressurizer temperature is 642 deg. F, but the temperature in the
- down s t r e am tail pipe to the PORV only ir.ditates 280 deg. F and quench tank pressure is 50 psi. Explain why these indications are accurate.
i
! ANSWER 5.09 (1.00)
- The steam leaving the pressurizer is at saturation conditions. Therefore, j it expands i se n thal p i c al l y across the PORV to the saturation conditions of the-quench tank. At 50 psi, the saturation temperature is about 280 deg.F.
REFERENCE FCS 007-007-08 193004K115 ...(KA'S)
(*+++* CATEGORY 05 CONTINUED CN NEXT PAGE +++++)
So_hBEORX._DE_tduCLEGE EQUEE_ELAUI_QEEEaIlOU,_ELulDS,_.atdD PAGE 7 IBEEMODYNed1CS QUESTION 5.19 (1.50)
Why are you mc. ,-
likely to form voids in the reactor vessel head during natural circulation?
ANSWER 5.10 (1.50)
The vessel head cool s much more sl owl y than the rest of the RCS during natural circulation because there i s very little circulation in the head area. (0.5) As the bulk of the RCS is cooled and depressurized, a saturation condition will exist in the head when RCS pressure reaches the saturation level for the temperature in the head. (0.5) At this point void begin to form in the vessel head area. (0.5)
REFERENCE FCS 007-007-16 193003K102 ...(KA'S)
QUESTION 5.11 (3.00)
Indicate the change in DNBR (INCREASE, DECREASE, or NO CHANGE) for tha f ol l owi ng pl an t transients:
- b. Pressurizer spray valve sticks open while operating at 100% power
- c. Reactor power is reduced frem 100% to 50% and Tavo is maintained at program value
- d. Letdown is i sol a t ed.
ANSWER 5.11 (3.00)
- a. decrease
- b. decrease
- c. increase
- d. no change (0.75 es)
REFERENCE FCS 007-Ob7-19 19300SK105 ...(KA'S)
(***++ CATEGORY 05 TONTINUED CT--I NEXT PAGE *++++)
. 5.__IBEGEY_DE_UUCLEa2_EOWER_EL6NI_OEERGI1CU _ELU1DS._6SD PAGE 8 IBEEdQDYNAMlCS
' QUESTION' 5.12 -(1.00)
If one of two centrifugal pumps operating in parallel trips, the system flow rate will go to (choose one of the f ol l owi ng) LESS THAN~1/2, ABOUT i 1/2, or MORE THAN 1/2 of its original value. Assume'the configuration and condition of all remaining system components do'not change at the time of the pump trip.
ANSWER 5.12 (1.00)
MORE THAN 1/2 (1.0) ;
REFERENCE t FCS 007-007-21 191004K109 ...(KA'E)
GUESTION 5.13 (1.50)
Describe how indicated reactor power will compare with actual reactor power when N1 adjustments are made using cal or ime tr i c data wi th the f ol l ow i ng I errors. Consider each case separately.
- a. Ac tual feed temperature is less than indicated feed temperature.
( b. Steam quality is less than the value assumed far the calculation.
ANSW iR 5.13 (1.50)
- a. Indicated power will be ( ac tu al power. (nonconservative) (0.75)
- 6. Indi cated power will be > ac tual power. (conservative) (0.75) r REFERENCE FCS 007-007-24 -
193007K106 ...(KA'S) i t
l QUESTION 5.14 (1.00) ,
i HOW and WHY will s te am gener ator pressure will vary for a constant Tave as '
an increasing number of tubes are plugged.
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---yge ,,,e ys w- ,-..-,.,,,-vy,,,.v...,-,,,,,-,,-.w- ,--r- Me-w ri rre - - gmx-ee -m = - m + w -c-- w ------u-.-+'r-o- w m w erew+m"
' 3 _'IBEDEY_DE_UUCLEAR_EDWEE_ELAUI_DEERATIDU4_ELUIDS4_eUD FAGE 9 IHEEMDDXUAMICS ANSWER 5.14 (1.00)
Steam generator pressure will decrease (0.5) because the total heat transfer area decreases as tubes are plugged. (0,5)
(Q = UA (Tave - Tstm) "A" decreases while "U" and Tave remain constant.
Therefore, for "Q" to remain constant, Tstm must decrease. Since the secondary side is at saturation conditions, Pstm must also decrease.)
REFERENCE FCS 007-007-18 193005K103 ...(KA'S)
QUESTION 5.15 (2.50)
True/ False
- a. Cr i t i c al rod height is dependent on how fast control rods are withdrawn.
- b. Critical rod height dictates the r e ac tor power level when criticality is achieved.
- c. The FASTER the approach to criticality, the LOtJER the reactor power level Will be when criticality is reached,
- d. While in the source range, changes in reactor power do not atfect Tavg.
ANSWER 5.15 (2.50)
- a. f al se
- b. false C. true
- d. true (0.625 ea)
REFERENCE FCS 007-005-04 192008K114 ...(K4'S)
(+++++ CATEGORY 05 CCNTINUED CN NEXT F4GE +++++)
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5.__IBEDEY_DE_UUCLEeE_ECWER_EL&NI_DEEEeI10N # _ELulDS,_6NQ PAGE 10 ,
IbEEMQQYuad1CS I i
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GUESTION 5.16 (1.00) l Descr i be 9;hy the KW/FT monitoring limit and the DNBR monitoring limit for ASI are more restrictive in_the negative direction than in the positive
< direction.
I '.
ANSWER 5.16 /.1.00)
More severe conditions exist toward the top of the core. Higher temp. and l owe r pressure. (1.0) ,
REFERENCE !
t FCS LP 7-5-6 EO 6.2 191004K112 ...(KA'S) '
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(****+ END OF CATEGORY 05 +++++) I r
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6 l P1 Mh? S'fSIEUS_DESIGU.9. CQUIEQL,_aSQ_1USIEUMEUIGIlCU PAGE 11 4
, GUESTION 6.01 (2.00) d Identif/ the automatic. system respon:e of ti.e Centainment Air Cooling and 1 Fil tering System f oll owing a large break LOCA inside containment. Assume <
normal system line up for power operations prior to the event.
.i i t 1
ANSWER 6.01 (2.00) '
a L
r Standby fan starts (all four running) (0.5) ;
}- CCW f l ow .to c ool er s is increased (valves open) (U 5)
L air flow is redirected through the HEPA and chercoal filters (;nlet and i outlet dampers open, bypass damper, shuti (0.5)
As temperature rises, plenum discha ge hatches operate to redistribute l exhaust duct flow. (0.5) l T !
I REFERENCE- .
j FCS LP 7-14-2 pgs 14 - 16, CBJ. 1.6
[
022000K402 ...(KA'S) j i
i 'i i QUESTION 6.02 (2.50) l l With the Steam Generator Blowdown System in a normal line up for 100% power l operation, answer the followingt f a. Hcw will the bl owdown tystem respond to high radie+ ion levels (above '
j al arm se tpoi n ts) due to a tube leak? Ge specific. s0.5) t i
- b. Describe the release path that exists through the L'owdown system when -
it is l i ned up r.ormal l y. Is release via this path nonitored? (1.0) l
- c. Hcw should the bl owdown system be realigned to handl the contamisated water. (1.0)
4
' i _.'ELedI_SYSIEUS_DESlE& _CDUIEDL eMD_INSIBUMEUIGIlDU PAGE 12 i ,
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t ANSWER. 6.02 (2.50) i i )
I
- a. Blowdown i sol a t i on val ve s HCV-1387A/B, 1388A/B shut. (0,5) l
- i. t j
- b. Water entering the blowdown tank flashes to steam. Since the tank is j continuously vented to' atmosphere, some of the steam and non ;
condensable gases are released to the environment. (0.5) This path is l not monitored.'<0.5) OR Water f rom bl owdown tank pumped by transfer
!" pumps to RW header which is monitored. .
j c. Isolate discharge to Raw Water (0.5) and line up to Waste Disposal !
System. (0,5)
, 000037A113 ...(KA'S/
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QUEST 70N 6.03 (1.00) !
l . ;
l Why is it important to reset PPLS and CPHS l ockou t relays BEFORE resetting SIAS lockout relays? l l
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! ANSWER 6.03 (1.00) 6 I
l The SIAS lockout relay coil will "burn out" if it is reset before clearing the in i t iating signal s (PPLS, CPHS). (1.0) ;
l l L
j REFERENCE FCS LP 7-12-5 EO 3.04 .
! 013000K401 ...(KA'S) .
I i
OUESTION 6.04 (2.00) !
l' Briefly describe the design of ti Reactor Vessel Level Indicating system ;
! and HOW it generates a vessel le>el signet. ,
4 1 r
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.- _ . - - - - . . _ - - . . . - - . _ . = . - - - - -. - _ _ _ - -
L_IE L ANI_SYSIEUS_ DESIST 4_CDUIRDL._ AN D_luSIRUMEUIGIl DU PAGE 13 f
I ANSWER 6.04 (2.00) !
The RVLI system uses strings of paired thermocouples spaced at prescribed i q
in terval s and arranged ver t icall y from the vessel head to the top c4 the active fuel. (1.0) '
One of the thermocouples in each pair is heated electrically while the other is not. As long as there is water surrounding a thermocouple pair the temperature difference between the two is small. However, when the water level drops bel ow a thermocouple pair the temperature difference rises significantly due to the lower heat transfer capability of steam.
This differential temperature variation is used to establish vessel level.
(1.0)
REFERENCE FCS LP 7-12-23 E0 1.04
- 002000K402 ...(KA'S) l l
QUESTION 6.05 (2.00) I i
! a- What provides backup cooling capability to bring the plant to a safe
! shutdown condition f Component Cooling Water is not operable? (1.0) l b. What is the potential public hazard associated with the use of this !
i CCW backup system even if t:,ere is. no loss of RCS i n te gr i ty? (1.0) i i
i h c ANSWER 6.05 (2.00) ;
- a. Raw Water (1.0) (supplied to various CCW cooled components ,
individually) !
- b. Failure to properly isolate the CCW system prior to initiating FW woul d rel ease toxic chemicals into the river. (1.0)
REFERENCE i e
- 0 000 26A10 3 ...(KA'S) i DUESTICN 6.06 (1.00)
I How is the operation of the R!161 affected by ; lacing the nornial/i nver s i on switch in INVERSION? When is this position used) {
f j i i ,
- - - . . - . - . - - . . . _ . - . - - . - . - . . - - - - - - . - . - - . - - ~ . - . - _ - -
i ' A_ 'EL6MI_SYSIEldS_DESISU _CDUIEDL.. 4 eUD_IUSIE1?.dEUIeI1DU PAGE 14 :
i- i
- i 3
J i i ANSWER 6.06 (1.00) '
t
! Placing the switch in INVERSION increases both the alert and al arm ;
j setpoints by a fixed amount. (0.5) It is used when a radon inversion j
! (temperature inversion) has been confirmed since the radon will add to the l j background level' increasing the possibility of erroneous alarms and !
a component activat,on. (0.5) ;
I REFERENCE i I FCS LP 7-12-3 EO 6.2 ;
073000K101 ...(KA's) !
j QUESTION 6.07 (2.00) lt t
Give two conditions which will cause automatic operation of the
)
a.
electric fire. pump. (1.0)
- t
- b. Give two conditions which will cause automatic operation of the diesel fire pump. (1.0) !
1 2 >
i (2.00) i 4
At4SWER 6.07 i
- a. 1. Header pressure dropping to 109 psi. (0,5) I
- 2. Whenever any transformer spray temperature bulb calls for deluge valve to open. (0.5) i '
j b. 1. Header pressure dropping to 99 psi. (0.5) 7
- 2. When the electric pump fails to restore pressure (or fails to ;
start) within 10 seconds of receipt of an auto start signal. f (0.5) l
! REFERENCE l FCS LP 7-11-12 ,
1 066000A202 086000A301 ...(K4'S) i i I :
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(*++++ CATEGORY 06 CONTINUED ON NEXT PAGE +****) ;
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l 6._IELGUI_SYSIEMS DESISU _CDUIEDL4_AUD_IUSIEUMEUIGIIDU PAGE 15 j QUESTION 6.08 (2.50T
- a. What conditions will generate an AFAS? (1.0)
- b. After automatic initiation, what generator level will the auxiliary feedwater system maintain without operator intervention? (0.5)
- c. In order for the operator to control SG 1evels from CB-10 using aux feed after AFAS initiation, what is the required position of the switches for HCVs 1107A/1108A and HCVs 11078/11088 on Al-66A/B? (1.0)
ANSWER 6.08 (2.50)
Steam press < 500 psi: <32% WR level to SG w/75 psi highsr than other (0,5)
- b. Level will cycle between 32% and 60% WR. (At 60% WR the AFA5 is automatically reset causing HCVs 1107A/B and 110eA/e to snu;. AFAS re-initiates at 32% WR.) (0.5)
C. 1107/SA OPEN (0.b>
1107/8B CLOSE (0,5)
REFERENCE FCS LF 7-11-1 E0 1.2 1.4 013000K107 . . . ( KA ' S )
GUESTION 6.09 (1.50)
At 100% power, F+edaater Regulating System in autcmstic, a high downcomer level occurs on the 2A S/G. Assume no operator ectier.
- a. What actions wi 1 occur with the FRS for S/G 2A whii - the higt downcemer level ;ignal is input? (1.0)
- b. What will occur v, th the sama system when the signal clees? (0.5)
I
-d._IEL6UI_SX5IEUS_ DESIGN._CQUIBOL+_6MD_1USIEUUEUI6110U PAGE 16 l
ANSWER 6.09 (1.50) ,
l
- a. Controller shifts to manual (signal locks in) (0.5) and the FRV close.
- (0.5)
, b. "Lock in" manual signal is input when the open signal clears causing the valve to open to the pre-high level condition position. (0.5)
REFERENCE FCS LP 7-12-15 EO 2.3 059000K104 ...(KA'S)
OUESTION 6.10 (2.50)
In going from source range to full power, the WR NIS employs two swaps (changes in its operation) to provide continuous power. indication.
- a. Describe the nature of each swap and the power level at which it '
occurs. (2.0)
I
- b. What will be the indication seen by the operator as each swap occurs?
i (0.5) i i
ANSWER 6.10 (2.50) ;
- a. At about 1000 CPS /4x10E-5% (0.5), the Ni$ goes from CPS to % power (0.25) and shifts from 2 detectors to 1 detector input (0.25). .
At about 10E-2% power (0.5), the NIS shifts from a log count rate circuit to a campbelling circuit (0.5).
F
- b. For the first swap, the operator will ,e e the CPS i ight go off and the #
j % power light come on at CG-4. (The chart recorder on CB-4 will show a drop of 1-2 decades when the swap occurs.) (0.25)
For the second swap, the operstor cannot readily identify that the swap has taken p l ac e . (0.25) ;
i i REFERENCE .
! 015000A103 ...(KA'E) 4 r 1 I e
i 1 ;
! (*++++ CATEG0F( 06 CONTINUED ON NEXT PAGE +++++) i l !
1 i
i ;
- 6. _._ E LAUI_ EM SI Etd S _ D E S IGN . __ C CUI E O L._6d D _1U SI E UU EUI6Il OU PAGE 17 GUESTION 6.11 (1.50)
- 3. Excluding the test s uai t c h , what are two (2) conditions that will cause an emergency diesel generator to automatically start and accelerate to full speed? (1,0)
- b. How are the suto start capabilities of an EDG affected by pl ac i ng the 183 tiaster Emergency Ewitch in the emergency mode at AI-133 M B' (0.5)
ANSWER 6.11 (1.50)
- 3. OPLS (0.5), Bus undervoltage (27 relay) (0.5)
- b. The auto starts are blocked. (0.5)
REFEPENCE FCS LP 7-13-5 EO 1.16, 1.9 064000A301 ...(KA'S)
OUESTICc! z.12 (2.00)
- 3. What provides indication of RCE l e' e l /v ol um" in the control room when the svetem i= drained below the piessurizer level taps' (1.0)
^
- b. Aside f r om indication, what are two speci+ic usem o/ the signals -
generated b ;. '
.ie Ica range t r e s s u r i :: a r p r e = =. ra detector' (1.0)
ANI4J E R 6.12 ( 2 . 0 0 '>
- a. Control room indication is p r c'> i de d b5 dif+eren'i 91 pressure transmitter (LI 1 ; '1 t2 tween a low presturire' vsp and a hot leg.
- b. 1. Shutc:an ccolin; tressure interlock. (0.5
- 2. .. c o temperatur - cuerpressura protec* ion, t 5=
REFERENCE FC- LP 23, 22, 20, $0 I-4 002000E410 ...(FA'I r ++++> CATEGORI ,
C' C T aT I'!UE C W t .E '" FmGE +++++
-m
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'6 __ELeNI_SYSIENS_ DESIGN +_CQUIROLo_6MQ_lNSIEUNENIellON PAGE 18
]- i j'
l i QUESTION 6.13 (2.50) t
- .s. What three (3) conditions can cause a Rod Block? {
-b. How can the operator reposition CEDMs as necassary to clear a Rod Block?
l
- c. What two (2) conditions / signals can cause a Rod Withdrawal Prohibit? t i
+ l' I
ANSWER 6.13 (2.50)
- a. PDIL, Deviation, or Out ot Sequence / Overlap. (0.25-ea) ;
! b. By controlling the affected CEDMs in manual individual while holding (
l the Rod Block Bypass Switch in "bypass". (0.75) i i i
, c. Hi p owe r: pre-trip (0.5) !
l Hi SUR pre-trip (0.5) t i i I REFERENCE !
j 001050K401 ...(KA'S) ;
l t l
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! i f I i' i l C i
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2._IEROCEDUEES_=_UQEdaL,_GEUGEdeL._EMEEGEUCY_auD PAGE 19 EaD10LOGLCaL COUIEDL i
OUESTION 7.01 (1.50)
- a. What are the normal OPPD permissible accumulated dose (PAD) limits as set by the Radiation Protection Manual? (1.0)
- 1. Daily
- 2. Weekly l-
- 3. Quarterly
- 4. Yearly
- 6. By what factor are the daily and weekly PADS increased during major maintenance or refueling projects? (0.5)
I ANSWER 7.01 (1.50) l
- a. 1. 100 mr
- 2. 300 mr
- 3. 1250 mr
- 4. 4500 mr (0.25 ea)
- b. 3 (three) (0.5)
[023;003H PEFERENCE FCS RPM Ull-2, LP 7-9-3 E0 1.3 194001K103 ...(KA'S)
OUEST19N 7.02 (1.50)
A Very High Radiation field exists in an area which cannot be controlled by locked acceso. What three (3) al ternate measures are required by Technical Specification section 5 to control access?
5 i
.2 __EROCEDURES_= NDEtML._eBUQEtML ._EtiEEGENCLAND PAGE 20 ,
RaQ10 LOGICAL COUIROL i
I ANSt4ER 7.02 (1.50)
- The area shall bei
- 1. roped off such that an individual at the rope boundary is exposed to
(= 1000 mrem /hr.
- 2. conspicuously posted.
- 3. a flashing light shall be activated as a warning device.
(0.5 ea)
REFER &8CE FCS TS 5.12.2, LP 7-9-3 EO 2.0-194001K103 ...(KA'S).
i GUESTICN 7.03 (1.00) 1 The 161KV system becomes. inoperable while the plant is at 100'/. power and is expected to remain so for at least 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. What actions are directed by Technical Specifications to be performed immediately?
l ANSWER 7.03 (1.00) 1.
Verify remaining power source
- 2. Notify NRC (by telephone or telegraph)
(0.5 each)
, REFERENCE FCS LP 7-13-1 EO 1.8
! 062000SG3 ...(KA's) i GUESTION 7.04 (2.00)
, OI-VA-2 provides guidance on how to isolate and identtfy a leak in the Aux
' Briefly describe these building which caused a Stack High Radiation Alarm.
2 actions.
4 5
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, ,.r,. ,n --,- - ,
I-
, 24_.' REDCED11EES_.:_NDEdeb_GENDReak_ FM F RGEUCLauD PAGE 21 RGDIDLDSICAL_CDUIRDL t
i-
) ANSWER 7.04 (2.00) l 1. Secure supply and exhaust fans l' 2. I sol ate all rooms
! 3. Start ONE exhaust fan
! 4. open dampers one room at a time until the leak is found
' (area rad monitors may also aid in leak i sol a t i on)
(0.5 ea) l REFERENCE
! QUESTION 7.05 (2.00) l l According to AOP-22, "Reactor Coolant Leak," explain four (4) way to verify l
that the code safeties or PORVs are not responsible for the loss of RCS i inventory.
t i ANSWER 7.05 (2.00) l (ANY 4 AT 0.5 EA)
- 1. Normal quench tank temp
- 2. Normal quench tank press
- 3. Normal quench tank level
- 4. No relief header temp alarms
- 5. Nn relief header acoustic flow alarms REFERENs FCS LP 7-17-22 EO 1.0, AOP-22 002000K405 ...(KA'S)
GUEST 10N 7.06 (1.00)
What is the operator's first action to attempt to terminate an uncon tr ol l ed CEA withdrawal?
ANSblER 7.06 (1.00)
Place the CEA drive control selector switch to "OFF" position. (1,0)
l 2._2EEDCEDURES_ _UDEtjeLo_ABUDEMe4_EMEEGENCY_AUD PAGE- 22 '
EADlDLDDiceL CDUIEDL' r
REFERENCE ,
FCS LP 7-17-27 EO 1.1, AOP-27 000001K301 ...(kA'S)
GUESTION 7.07 (2.00)
Per EOP-01, how _is.the RCS heat removal safety function verified (two answers required)?
AN SW ER - 7.07 (2.00)
- 1. Verify at least on S/G has level:
- a. 35-85% NR with feedwater available (0.5) i b. Being restored by adequate feed flow.-(0,5)
- 2. Verify steam and bypass valve functioning to controls
- a. RCS to 525-535 Deg.F (0.5)
- b. S/G pressure 850-925 psia <0.5)
REFERENCE FCS LP 7-18-1 EO 1.3, ECP-01 000007KS01 002000K111 ...(KA'S)
QUESTION 7.08 (1.50)
If reactivity control is not established per EOP-1, "Reactor Trip," what two (2) contingency ac t i ons mi. it be taken by the operator?
4 ANSWER 7.08 (1.50)
- 1. Open CEDM clutch power supply breakers. on pane) AI-57 (0.75)
- 2. Emergency borate (0.75)
REFERENCE FCS LP 7-18-1 EC 1.4, EGP-1 000024K302 ...(KA'S) 4 1
1
(****+ CATE9ORY 07 CONTINUED CN NEXT PAGE *++++)
i 4
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2.__E20CEDUEES_= JUDET $aL._oEUGEMeL,_EMEEGEUCY_GUD PAGE. 23 ,
EADIOLOGICAL_CQUIEQL
-QUESTION 7.09 (2.00)
During Reactor Trip Recovery, the operator is cautioned not to overreact to
- low S/G level and add f ee dwater sl owl y to avoid problems. What are four
.<4) of the potential problems to be avoided during recovery of S/G level?
j i I
i J
1 ANSWER 7.09 (2.00) i
- 1. Excessive cooldown
- 2. Excessive transient in pres.aurizer level and pressure
- 3. Overfilling S/Gs
- 4. Possibili.ty of feed ring damage / water hammer if feeding via feed ting i and S/G l evel drops bel ow f eed r i ng.
(0.5 ea) .,
1 REFERENCE i FCS LP 7-18-1, EO 1.4, ECP-1 000007K106 000007K301 ...(KA's) l t
I QUESTION 7.10 (1.00) ;
i Explain why the operator is directed to cool down below 510 deg.F prior to ,
l ANSWER 7.10 (1.00)
=
l i Reducing RCS Th to <= 510 deg.F will maintain adequate RCP NPSH and RCS l subcooling when RCS pressure is reduced below S/G safety valve setpoint.
- (1.0) 4 ,
i REFERENCE ['
l FCS LP 7-18-4 E0 1.12, E0P-04
- 000038K306 ...(KA'S) l e
- f OUESTION 7.11 (2.00)
List four (4) of the actions an operatcr must take if the reactor becomes i j critical be l ow the ECP whiri performing OP-7. [
L
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- _ - .--..., -.-- - . ~
2.a_lEEDCEDUEES_ _UDEUGLo_aEUDEt%- FMFRGENCY_GUD PAGE 24 EADlDLDGICAL_CDUIBOL ANSWER 7.11 (2.00)
(ANY 4 AT 0.5 EA) j 1.
Make the reactor subcritical (insert CEAs)
- 2. Flace all 4 Extended Range Cutout switches in "off"
- 3. Notify the Reactor Engineer (or designee)
- 4. Verify boron concentration, time and Tavg
- 5. Recalculate the ECP
- 6. With Reactor Engineer's permission, restart the reactor, otherwise go to hot shutdown and notify Manager-FCS (or designee)
REFERENCE FCS LP 7-51-2, E0 1.3, OP-7 001000SG8 ...(KA's)
QUESTION 7.12 (1.50)
For questions a. through c., select the best answer for each question from the list 1. through 4. below. Some answers may be used more than once or not at all.
Select the best answer:
- 1. 15 min
- 2. 30 min
- 3. 60 min
- 4. a shift
- a. If any of the recorders itemi ed in the Waste Liquid Releas.e Operations Checklist are inoperable, at what interval must applicable date be recorded in the Liquid Discharge log?
- b. If any of the recorders itemized in the Containment Purge Release Operations Checklist are inoperable, at what interval must applicable data be recorded in the Containment Purge Release log?
- c. How long shall each monitor tank be recirculated prior to s.ampling?
ANSWER 7.12 (1.50)
- a. 2. 30 min
- b. 3. 60 min
- c. 2. 30 min (0.5 ea)
, _ . . _ _ _ . .__-.____.__ _ _ _____ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ ~ _ _ . - _ _ _ . _ . . _
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' 2._ '.EEDCEDUEES_ _UDEdet._eBUDhtdeL 4_EUERDENCLeND PAGE 25 l BeDlDLDDlCGL_CDUIBOL I !
REFERENCE '
i FCS LP 7-15-4 E0 1.8, 2.7, S.0. T-2 }
l 028000SG7 068000SG3 068000SG7 ...(KA'S) l 1 i OUESTION 7.13 (2.00) l l.
i '
- a. What is the minimum time delay atter power operations before handling of irradiated fuel can begin? (0.5) ;
I
- b. When irradiated tuel is moved to the Spent Fuel Pool, where must it be j placed in the pool? (0,5) '
k c. List two (?) evolutions for which Base Count Correction Data Records l must be completed per OP-11. (1.0)
I i l i i !
I '
1
! ANSWER 7.13 (2.00) }
I l a. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (0,5) l I !
! b. Region I (borated or poisoned regton) (0.5) {
- i i
i c. (any 2 at 0.5 ea) l 1. Transfer et a source from one assembly to another i 2. Movement of an assembl y containing a source ;
- 3. Base count corrections may be perfromed as new fuel is moved [
l
- around a source. :
r i
I ,
i REFERENCE
! FCS LP 7-11-13 EO 2.1, 2.11, OP-11 i j 034000A401 034000SG12 034000SG7 ...(RA's) t OUESTICN 7.14 (2.00) l Briefly describe three (3) of the alternate methods listed in AOP-19 for [
residual heat remeval on loss of Shutdown Cooling. Assume normel SDC f cannot be restarted. L l
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2o_lEEDCED1JEES_ _UDEdeL_GEUDEtk6_EL1EEDEUCX 6UD PAGE 26 l EADlDLDDLCAL-.CDUTEDL , .
I 2
i ANSWER 7.14 (2.00) 1 i' (ANY 3 AT 0.5 EA)
- 1. Es t abi l sh Chargi ng and Le t down
- 2. With head installed, line up f or natural /f orced circual tion using SGs.
- 3. Estabilsh f l ow through SDC suction through containment spray - HPSI -
through loop injection valves
- 5. Head removed, Fuel transfer canal is open and refueling pit is flooded, align fuel transfer canal drain pumps to take a suction from the SIRWT to SFP and align storage pool pumps to take a suction from the SFP to the SIRWT. (Ensure suction is not taken from the transfer canal drain.)
i REFERENCE l FCS LP 7-17-19 E0 1.0, AOP-19
, 005000SG10 ...<KA'S) 1 1
i OUESTION 7.15 (2.00)
For each item, answer TRUE or FALSE concerning 01-DG-2, "Normal Operation, Diesel Generator No. 2":
- a. Diesel Generators may be tested cencurrently if one diesel is maintained running for one hour after shutdown of the other.
- b. The Diesel Generator can be loaded when Jacket water temperature reaches 100 deg.F.
- c. Placing the master emergency- s-witch (183/mes) at AI-183A/B out of the normal position will transfer control of the diesel to the local control panel,
- d. A diesel generator abould not be started or loaded in anticipation of a loss of offsite p ow e r .
- 49 DER 7.15 (2.00)
- a. false
- b. false
- c. true
- d. true
l l . .
2o_ERDCEDUEES_ _UDR!'ALo eEUDEtdaL_EliFRSEUEY_GUD PAGE 27
, RADIDLDD1ChL_CDUIBDL i
i 1
l REFERENCE l FCS LP 7-23-5 EO 0.57, 01-DG-2 l 064000K102 064000K303 064000K401 ...(KA'S) 7 l
i n
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p i
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. . . . . . _ - . _ , _ _ _ . . _ _ - . _ _ . _ _ - . _ _ _ . . _ m._. _ . _ _ _ _ . _ . . _ - . _ _ .
'Bo-lhDdlUlST2aTlUE 2BDCEDUBESo CDUD111DUS. eUD LlulIhIlDUS PAGE 28
. t b i
) I QUESTION 8.01 (1.50) ;
4 i,'
j What three (3) conditions must be met to allow termination ot decay heat j j removal for up to eight hours per T.S. 2.1. ;
1 r i
ANSWER G.01 (1.50)
- 1. no dilution
- 2. no refueling operations
- 3. all containment penetrations closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> ;
! (0.3 ea) j REFERENCE FCS LP 7-11-23 E0 1.7, TS 2.1.1 '
j 005000K307 005000K509 ...(KA'S) :
c i s
- i i GUESTION 8.02 (2.50) i l The f o l l ow i n g r e l a t e to Hol d and Cau t ion Tag requirements.
i ,
l a. Who will request the hanging and clearing of " Hol d Or der " tags? (0,5) j i
- b. According to Standing Order 0-20, in what two (2) si tuat i ons will "Temporary Clearance" tags be used? (1.0) !
l [
! c. Give two (2) general situations / conditions in which "Caution" tags are l
! used. (1.0) ;
I .
1 i a 6 l l-l 1 ANSWER 8.02 (2.50) !
I t a. Sys tem operator (system operations /906 personnel) (0,5)
- b. 1. When equipment is to be operated 5n a test position or mode (such as trip checks, motor rotation checks, limiterque operator
[
settings, etc.) (0,5)
! 2. When necessary to allow system / equipment usage for unusual !
circumstances. (0,5) j {
- c. 1. When special instructions in the operation of ccmponents must be i exercised. (0.5) i
- 2. During system turnover to specity components ano systems under i OPPD control. (0.5) i I
e I
1 I
I B.__aDtd1NISIRGI1UE RRDfEDURRS 4 _CDUDlIlDNS 4 _aND_Liti1IaIIDUS PAGE 29 REFERENCE FCS LP 7-15-3 EO 1.1, 5.1, S.O. 0-20 QUESTION 8.03 (2.00)'
- a. An i r.d i v i du a l receiving a bomb threat at FCS should immediately report the-occurrence to which two (2) personnel (by title)? (1,0)
- b. List two (2) agencies outside of FCS/0 PPD which must be notified if a bomb threat is received per Standing Order G-16. -(1,0)
! ANSWER 8.03 (2.00) j a. Shift Supervisor (0.5) 3 Manager - FCS (0,5)
- b. (any 2 at 0.5 ea) l 1. Washington County Sheriff
- 2. Nebraska State Patrol
- 3. Bl ai r Fire Depar tmen t
. REFERENCE l FCS LP 7-15-4 EO 1.0, S.O. G-16 1
i i
i QUESTION 8.04 (1.50)
- a. What two (2) instances cause a "Flame Cutting and Welding Permit" to
- j. expire? (1.0)
- b. How often must a "Flame Cutting and Welding Permit" (Form FC-18) be renewed? (0,5) l l
l
, AN5WER 8.04 (1.50) l a. Completion work described in Maintenance or Design Change Order (0.5) l Two wee %s from date of issue (0.5) l b. Each thi4t (0,5) i j REFERENCE I FCS LP 7-15-5 EG 1.7, 1.9, S.O. M-9 l
(+++++ CATEGOR1 09 CONTINUED ON NE<T PAGE +++++>
l I.
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l Bu_ett!1NISIEeIl!.!E_EEDCED1)RES4_CDNDlIIDUS _eUD._LltdileIICUS PAGE 30 )
i
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i QUESTION 8.05 ( .50) ,
If the Shift Supervisor was not the originator of an Operations Incident .j (01) report (Form FC-79), what does'his signature on the 01 signify? !
i AN SWE R ' 8.05 ( .50)
That he has been informed about-the incident being reported. (0.5)
REFERENCE i FCS.LP 7-15-5 EO 0.1, S.O. R-4 '
b OUESTION 8.06 (2.00)
}
} a. When initially notifying the NRC Operations Center of a significant
' even under S.0. R-11, what two (2) pieces of information should be transmitted? (1.0) j b. When does the "clock" start for events reportable to the NRC within l one hour? (1.0) i I
l l ANSWER 8.06 (2.00) i
- a. 1. The emergency classification (0,5)
- 9. . If the report is a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report (0.5)
- 6. Uhen the euent occurs or is recognized (EAL declared) (1.0; REFERENCE FCS LP 7-15-5 EO 3.18 S.0. R-11, LP 7-51-6 EO 1.0
(***** CATEGORY 08 CONTINUED ON NEXT PAGE +++**)
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Bo__aDt11UISIEeI1UE_EEDCED1)EES _CDUD1IIDUS, auD_LitilI6IIDUS PAGE 31 l 1
i I
i i GUESTION 8.07 (2.00) 1 j~ !
j a. The Official Control Room Log ts consider to be the most important log )
j of the Operations staff because: - (select the best answer) (0.5). i i 1. It is a legal document :
- 2. .I t de tail s any plant problems j
- 3. It is the most up-to-date log !
j
- 4. It provides a complete description of plant activities. j f b. Which two (2) individuals (by title) are responsible for maintaining the Official Control Room Log? (1.0) {
! c. How often must the Shift Supervisor review and sign the Official ,
} Con t r ol Room Log? ,
(0,5) :
1 I
d i.
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ANSWER 6.07 (2.00) [
- a. 1 (0.5)
- c. at the end of each shift (0.5) !
i REFERENCE .
l FCS LP 7-15-7 EO 4.3, 4.4, S.O. O-24 i i
k
! OUESTION 8.08 (1.50) f
- i
! List three (3) components which must be operable to monitor leakage into l l containment per T.S. 2.1.4 !
a i 1 l I ;
i I
- AN94ER 8.08 (1.50) ,
I !
1
' (ANY 3 AT 0.5 EA) r r
)
- 1. Containment dew coint instrument t
! 2. Containment radiation monitor I
- 3. Containment sump level instrument i
, REFERENCE I FCS LP 7-51-6 EO 0.4, EOP-2 .
002000A201 !
3 000061K302 ...<kA'S) i s t i i
- i
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i B.__.6DUlNISIEGI1UE_EROCEDUBES4_EDUD1IJDUS _GUD LIM 1IeIIDMS PAGE 32 a
i
! QUESTION 8.09 (2.50) i'
- a. Why do Technic al Specifications require di sabl i ng HPSI pumps when Tc. !
i s bel ow cer tai n val ue s? (1.0) {
l
! b. List three (3) condi t i ons wh i ch woul d al l ow one or more of the HPSI
! pumps to remain operable even when Tc 'i s b e l ow the temperature l
{ requiring disabling of the HPSI pumps. (1.5) l l ,
I i
l ANSWER 8.09 (2.50) !
t !
j a. Prevent inadvertent actuation of HPSI which could overpressurize the l j RCS (subjecting it to brittle fracture conditions). (1.0) {
- b. (any three at 0.5 each) l I
I
- 1. Reactor vessel head removed
- 2. Pressurizer safety valve removed
- 3. PORV removed ,
4 Charging pumps are inoperable (HPSI may be used to inject boron) f 1
I, REFERENCt ;
FCS LP 7-51-6 EO 1.0, 1.2, T.S. 2.3 i 006050A201 006050K402 ...(KA'S) !
OUESTION 8.10 (3.00)
- a. Name the three (3) mai n barr i ers , u sed f or emergency level j classification criteria which protect the publ i c from the release of j fission products. (1.5)
Y
- b. Correlate barrier failure with three (5) different Emergency [
Classification Levels. (1.5) !
t 6
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I B GDMlulSIEGI1UE_ERDCED1)RES4 CDUDlIIDUS4 _eUD_Liti1IGIIDUS PAGE 33 f
- i l ANSWER 8.10 (3.00) ,
- a. 1. fuel clad l 2. RCS (Pressure boundary) -t
- 3. Reactor Containment i
- b. 1. Alert failure of one barrier !
i~
- 2. SAE failure of two barriers !
- 3. GE failure of three barriers l (0.5 ea) '
REFERENCE FCS LP 7-60-9 EO 0.4, 0.5, 0.8, EPIP OSC-1 i" ,
j QUESTION S.11 (3.00)
- a. What is the basis for the limits on radioactivity on the RCS coolan t?
! (1,0)
I
! b. Why are there provisions for allowing operation to continue when a
] particular sample is s i gn i f i c an t l y above one of the radioactivity l limits? (1.0) f l c. Why is there a RCS temperature limit imposed when a radioactivity j limit is exceeded? (1.0)
I ANSWER 8.11 (3.00)
- a. To limit the (2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) dose at the site boundary (0.5) following a SGTR. (0.5) l b. To accommodate the possib!v iodine spiking which may accompany changes
- in thermal powcr. (1,0)
- c. Prevents a release by keeping saturation temperature below setpoint for S/G atmospheric relief velves (main steam safety v alves) . (1.0)
REFERENCE 00 0 OK50- 0 1 00K 08 . . . < KA ' S ')
i
(*++++ CATEGORY 08 CONTINUED ON NEXT PAGE +++**)
i
i !
BA._ADt31NISIRaIIVE EEDCEDURES4 CDUD1IIDNS.,_AUD_ Lit 11IAIIDMS PAGE 34 ,
P ,
1 e >
t l QUESTION 8.12 (3.00) j 1- i
) According the S.O. G-28, Fire Protection Plan {
- a. Who is responsible for making initial _offsite notification? (0.5)
- 6. During backshift, who will be in charge at the scene? (0.5)
{ c. -What is the Shift Supervisor's responsibility during a fire on backshift? (0.5)
- d. IF a nuclear emergency develops during a fire, where should
! non-operating and non-security personnel assemble? (0.5) i Describe how the plant communications system is used during fire
, e.
l fighting activities. Also explain backup methods in the event of
- normal system failure. (1.0) l t
! ANSWER 8.12 (3.00)
! b. Licensed Equipment Operator (LO) (0.5)
- c. Take charge of the MCR operating crew. (0,5)
- d. Fort Calhoun store room (warehouse) (0.5)
- e. Gaitronics line i la used for fire fighting communications, (0.25) while line 2 is used for operations. (0.25) Sound powered phones (0.25) or portable rsdios (0.25) are used for backup.
I l REFERENCE
a 086000K202 ...(KA'S) a l
I I
l l
J
t U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSB EXAMINATION FACILITY: FT._LGLBOUb -_________
REsCTOR TYPE: _EUR CE__________________
DATE ADMINISTERED _SS4Q24Q2_.-- __ -,_______
EXAMINER: ._MCCSOEY._S._____________
CANDIDATE: ____ ________
luSIEUCI1OUS_IQ CAUDlDAIE Use separate paper for the' answers. Write answers on one side only.
Staple question sheet on top of the answer sheets. Points vor each ;
questien are indicated in parentheses atter 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) nours a.fter the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGOR(
__UALUE_ _IDIAL ___SCDEE___ _UeLUE__ ______________CGIESDEY_____________
_25.0C__ _25.00 ___________ ___ ___ 1. PRINCIPLES OF NUCLEAR POWER PL WT OPERAT I ON , THERMODYNAMICS, HEAT TRANSFER AND FLUlD FLOW 25.QQ__ _25.QQ ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SyGTEMS
_25.QQ__ _25.00 ___________
'7 INSTRUt1ENTS AND CONTROLS
_25.QQ__ _25.QQ ___________ ________ 4. PROCEDURES - NORMAL, AENORMAL, EMERGENCY AND RADIOLOGICAL CONTROL ,
i
- 100.Qa__ __ -- ________% Tot:Is Final Grade I
All work don? on this examination is my own. I have neither given
- nor raceived aid.
Candidate's Signature I
( i l
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, . ~ . . - . . _ _-_- --~- _.... - ._. - - _ _ _ _ . _ - . - - _ _ _ . _ . - - _ - . -
,l8 lo__BRidCIELES_DE_UUEthdR_EDWEE_BL6UI_DEEEAIIDU. PAGE 2 l 1HFRHnDXN6MlES.,_BEbI IE6USEEE bMD ELULD ELDW ;
i i s L ' QUESTION 1.01 (1.50) !
I i l~ With a constant 1 DPM startup rate, will the time it takes to change power j
- from 20% to 40% be LONGER, SAME, or SHORTER than the time it takes to go ;
from 60% to 80%? Use calculations to support your answer. ;
l l ANSWER 1.01' (1.50)
LONGER (0.5) 1 4
P=Po 10 E SUR(t) (0.5) l t= Log P/Po 4 t1= Log (40/20)= .3 min = 18 sec t2= Log (80/60)= .12 min = 7.5 sec (0.5 for proper use of equation)
REFERENCE FCS RXTH 007-005-04 192003K106 ...(KA'S)
QUESTION 1.02 (2.50)
- a. What is (are) the source (s) of decay heat following a reactor shutdown? (0.5)
- b. What factor (s) affect the magnitude of decay heat? (1.0)
- c. Why is RCS heat removal capability important with respect to decay heat? (0.5)
- d. Why is the power level represented by decay heat not indicated by the nuclear instruments? (0,5)
p a,, .
'NL -
.i j $1.+1 -
ERIUCIELES DE_UUCLFAR EDWER_ELAUI_DEEEaIIDW PAGE= 3 l y-Ag
- ,y IBEEMDDYNAMfC$ BEAT TRANC
- EER_auD_ELUID ELDU l N ' pn ' xt !
93 l
OY(it 3 __
h .* \
x kb s s 3 i(03OER
- 'u 1.02 (2.50) !
c At Decay heat i s produced f rom the decay ot f i ssi on produc ts ( f r a gmen t s) {
f ormed dur i ng power operation. (0.5) ;
.t b.. Power history (0.5) and time since shutdown (0.5) -
IO -
r
- c. To prevent fuel clad damage (failure) (0.5)
P d- The bulk of the decay heat is produced from fission products which do 4 not emit neutrons ( al pha, beta, or gamma emitters) as part of the decay process. (0.5) ,
REFERENCE
=FCS RXTH 007-005-04
- 192008K127 ...(KA'S)
QUESTION 1.03 (1.50)
Neutron flux increases with reactor power. Rod worth is partly a function !
of the flux to which it is expcued. If reactor power (and flux) is i
, increased un!formly throughout tue core, HOW will the worth of an 1 individual rod change and WHY?
i i
ANSWER 1.03 (1.50) I L
The worth of the rod will NOT change. (0,5) Rod worth variation is relative to the change in local flux ccapared to average +1ux. (CONCEPT) ,
(1.0) k
d i
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(***** CATEGORY 01 CONTINUED ON NEXT PAGE *++++) l
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1 5
- a r- .--.-..,-,--,.--.,+m-._,-,-,--,.,+-e-----m,,,mm,- -v,--.- , , - , - - - . - - -
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1 . . i j i._.ERIUCLELES_DE_UUCLEGE_EQUEE_ELAMI_QEEE6ILQN. PAGE 4 l j ~
IHERMCDYueU1CS._HE61_IRANSECE_6UD ELU1D_ELCU t
- 1 1 i d 1
}
l QUESTION 1.04 (1.50) ,
{- i 1 Assume the following conditions:
i
) -
BOC with a slightly positive MTC j -
Reactor power 16% j j
Steam dumps in manual ;
1 Turbine control in manual ;
i -
j A' peripheral rod drops during these conditions. Briefly describe and {
l explain the reactor response to this event assuming NO operator action. -
- I l !
I i
ANSWER 1.04 <!.50) i I i The dropped rod will cause reactor power to decrease (go suberitical) due !
to the negative reactivity insertion. (0.5) Since 6 team demand will remain !
constant, the RCS will begin to cooldown. (0.5) With a positive MTC, more negative reactivity will be added to the core as RCS temperature continues l1 to drop. (0,5) (Ultimately RCS temper ature will stabilize at the heat '
I addition capacity of the RCPs combined with a reduced steam rate based on j lower secondary temperature and pressure.) l I
I REFERENCE ,
FCS RXTH 007-005-03 !
172005K102 ...(KA'S) !
f GUEETION 1.05 (1.50) ;
t Give three (3) general actions which an operator may take to add negative I reactivit/ to the reactor while at power. Assume MOL core characteristics. f f
ANSWER 1.05 (1.50)
- 1. inser t rods
- 2. add bor on
- 3. reduce steam demand *
(0.5 each) j
("Increase Temperature" is not a valid response since the operator can i only affect temperature indirectly. Hoaever any valid action l described which results in a temperature increase may be accepted.)
l i
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. . _. - - .- .- - _ - . . - . , . - . - . . - ~ . . _
)
1.__ERINCIELES 3E_uuCLEGE_EL.R_ELAUI_QEE!:ellLG PAGE 5 ;
! IBEEMODYUaMICS _BEGI_IRANSEER_AND_ELU10_ELOW l t
REFERENCE l FCS RXTH 007-005-03
- l 001000K517 ...<KA's) j i
a QUESTION 1.06 (1.50) i i '
i Why do technical specifications impose both an upper and l owe r limit to the i magnitude of MTC? !
.j ANSWER 1.06 (1.50) ',
I 2
The upper limit (positive MTC) is set to ensure reactor stability. (0.5) j l
) The lower limit (large negative MTC) reduces the impact of a postulated !
steamline break with subsequent cooldown and potential for return to power.
(CONCEPT) (1.0) ;
i f 1 REFERENCE i
, FCS RXTH 007-005-03 192004K106 ...(KA'S) l 4-f -
y I l QUESTI ON 1.07 (1.50) !
l On a reactor trip f r om ecu l l i br i um Xenon conditions, how de the differential Xenon reactivity insertion / peak worth - equilibrium worth)
- and time'to peak intertion very as equ il ibrium condi t i ons are changed from !
100% to 20% power?
1 i
, ANSWER 1.07 (1.50)
The differentia, reactivity insertion decreases (C.75) as does the time to [
! peak inser t ien (0.75). l 2
RFFERENCE - :
- FCS RXTH 007-005-03 ['
j 000005K103 001000K538 ...(KA'S) a ;
f l OUESTION 1.09 (1.00> l
{
What is the basis for allowing a reduction in the n i r imum shutdown margin j requirement once RC5 temperature goes below 210 deg. F? l l l a
I I
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.L EBillCLELES-.DE NDCLEaB EDWEB-.EL6NT DEEBaTlDtk PAGE 6 1BFRMnD.YNeU1CS., BhhI TEANSEEB aND ELUlD ELDW
+
j_ ANSWER ~ 1.08 .(1.00) ;
The potential for a reactor restart due to a steam line rupture and cooldown'no !_onger exists below 210 deg. F. (1.0)
REFERENCE FCS RXTH 007-005-06
- 192002K114 ...(KA'S) >
l n
QUESTION' 1.09 (1.00) l The power operated relief valve-(PORV) opens while at oower. The pressurizer temperature is 642 ceg. F, but the temperature in the downstream tail pipe to the,PORV only indicates 280 deg. F-and' quench tank .
pressure is 50 psi. Explain why these indications are accurate.
' ANSWER 1.09 (1.00)
The steam leaving the p r e s s'J r i t e r is at saturation conditions. Therefore, ,
it expands i sen thal p i cal l y across the PORV to the sa t u r e.t i on conditions of the quench tank. At 50 psi, the saturation temperature is about 280 deg.F. ,
' REFERENCE FCS 007-007-08 193004K115 ...(KA'S) i t
QUESTION 1.10 (1.50)
Why are you more likely to torm coids in the reactor vessel head during naturai circulation? ;
i ANEWER 1.10 (1.50)
The vessel head cool s much more sl owl y than the rest of the RCS during ;
natural circulation because there is very little circulation i r, the head area. (0.5) As the bulk of the RCS is cool ed and depr e ssur i zed, a saturation condition will exist in the head when RCS pressure reaches the saturation level for the temperature in the head. (0.5) At this point void begin to f>rm in the vessel head area. (0.5)
1
- . ..=
.L_ERINDIELES_DE_NUDLEAE_EDWFP RtANT_DEERATION, .PAGE 7 IBEEMDDYNAMI DS_,_.BEGI_IEeNSEE R_ eUD_E LUI D_ E LDU REFERENCE FCS'007-007-16 193003K102 ...(KA'S)
OUESTION 1.11. (3.00)
Indicate.the change in DNBR-(INCREASE, DECREASE, or NO CHANGE) for the f ol l owi ng pl an t transients:
- b. Pressurizer spray valve' sticks open while operating at 100% power -
- c. Reactor power'is reduced from 100% to 50% and Tave is maintained at program value
- d. Letdown is i sol ated.
ANSWER 1.11 (3.00)
- a. decrease b, decrease
- c. increase
-d. no change *
(0.75 ea)
REFERENCE FCS 007-007-18 193008h105 ...(KA'S) f OUESTION l'.12 (1.00)
L If one of two cen tri f ugal pumps operating in parallel trips, the system e flow rate will go to (choose one et '. i . following) LESS THAN 1/2, ABOUT~
1/2, or MORE THAN 1/2 of its original value. Assume the configuration and l condition of all remaining system components do not change at the time of l
the pump trip.
ANSWER 1.12 (1.00) l MORE THAN 1/2 (1.0) l
ikwEElUCIELES_DE U1JCLEaR_2DWEE_EL.6NT_DEEEbIlDLL PAGE 6 IEEEdDDYNaL41CS4 _BEAI_IRANSEER_eUD_ELUID_ELOW REFERENCE FCS 007-007 '191004K109 ...(KA'S)'
QUESTION 1.13 (1.50)
Describe how indicated reactor power will compare with ac tu al ^re ac t or p ower When NI adjustments are made usi ng calorime tr i c data wi th the'following errors. Consider each case separately,
- a. Ac tu al feed temperature is less than indicated feed temperature,
- b. Steam quality is less than ' the value assumed for the calculation.
ANSWER 1.13 (1.50)
- a. Indicated power will be < actual power. (nonconservative) (0.75)
- b. Indicated power will be > actual power. (conservative) (0.75)
REFERENCE FCS 007-007-24 193007K106 ...(KA'S)
(
QUESTION 1.14 (1.00)
HOW and WHY will steam generator pressure will vary for a constant Tave as an increasing number of tubes are plugged.
ANSWER 1.14 (1.00)
Steam generator pressure will decrease ('v.5) because the total heat transfer area decreases as tubes are plugged. (0,5)
(0 = UA (Taue - Tstm) "A" decreases while "U" and Tave remain constant.
Therefore, for "Q" to remain constant, Tstm must decrease. Since the secondary side is at saturation concitions, Pstm must also decrease.)
REFERENCE FCS 007-007-18 193005K103 ...( A'S)
(+++++ CATEGORY 01 CONTINUED ON NEXT PAGE i++**)
1.__EELNCIELES_QE_UUCLEAE_ECWEE_ELAMI_QEEEAILQU. PAGE 9 IB ERM O D'fuAM ICS._B EAI_IEAU S EE R_AU D _ E LulD_E LCU OUESTION 1.15 (2.50)
True/ False
- a. Critical rod height is dependent on how fast control rods are withdrawn.
- b. Critical rod height dictates the reactor power level when criticality is achieved.
- c. The FASTER the approach to criticality, the LOWER the reactor power level will be when cr i tical i ty is reached.
- d. While in the source range, changes in reactor power do r.ot affect Taug.
ANSWER 1.15 (2.50)
- a. false
- b. f al se
- c. true
- d. true (0.625 ea)
REFERENCE FCS 007-005-04 19200SK114 ...(KA'S)
QUESTION 1.16 (1.00)
A cen tri f ugal pump discharging against ..... will draw the most motor amperage. (Select the best answer from below.)
- a. pa-allel pump head
- b. :e o back pressure
- c. a cl osed di scharge val ve
- d. the suction of a series pump ANSWER 1.16 (1.00)
- 8. zero back pressure (1.0) 1
1 - RElGCLELES DE UDELE6E EDUEE 2LGUI DEEEeIID% PpOE. 1o IEEEUDDYUaulCS BEeI IEaNSEEE abD_ELUID ELOW i
REFERENCE
.FCS 007-007-21 191004K112 ...(KA's) i r
I r
i I
\ l r
t l
)
l l
l t'
(*+++* END OF CATEGORY 01 ++++*) .
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'i _ELeMI_DESIGU M CLUDW G SeEEIX AND_ EMERGENCY _SYSIEUS PAGE .11 QUESTION 2.01 .(2.00)
Identify the' automatic system response of the Con t a i nmen t Ai r. Cool i ng and Filtering System following a large' break LOCA inside centainment. Assume ;
normal system line up for power operations prior to the event. ;
ANSWER 2.01 (2.00)
' Standby fan starts (all four running)'(0,5)
CCW flow 'o coolers is increased (valves spen) .0,5) (
' air. flow is. redirected through the HEPA and charcoal filters (inlet and outlet dampers open, bypass dampers shut) (0.5)
As temperature rises, plenum discharge hatches operate to redistribute exhaust duct flow. (0.5)
REFERENCE FCS LP 7-14-2 pgs 14 - 16, OBJ. 1.6 022000K402 ...(KA'S) '
QUESTION 2.02 (3.00)
Using figure 2-1, show and label all major penetrations into the Reactor Coolant System. DO NOT include instrument or test penetrations. (Eighteen properly placed and labelled penetrations are required for full credit.)
b l
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, . . . . - - . , . . _ . _ . . . , _ _ _ . _ _ _ . _ . . . , ~ , , . _ . . . . . . . . , _ , . ~ . . . . . _ , _ , . . - , _ . _ . , . . . _ _ . . _ . . . _ . _ , , _ , . . _ , , _ . . . . , , , , . . . . ~ . - _ . - . . . _
2a__ELeNI_ DESIGN _1UCLUQldG_SeEEIX_eUD_EUEEGEUC1_SYSIEUS PAGE 12-F ANSWER 2.02 (3.00)
/---RCP D---\ /---RCP B---\
'/ \<---S1 SI--->/ \
/ \ / \-->RCDT RCDT<--/ N VENT CHG-->/ \
/ \ /!\ / \
/ \ ! / . SURGE N y g____!________f ! g SG 28<-----!--------!-------! REACTOR !-----!----!----!------>SG 2A i i .i i i !
i i \ \!/ \!/ /-------------\ N!/ \!/ /
\ / \ /
\ /---> SPRAY \---> SPRAY /
RCDT(-!---\ / \ /-->RCDT
' N /(---51 SI--->\ /
LD(---! \---RCP C---/ \---RCP A---/
(EIGHTEEN PENETRATIONS AT 0.166 EA) :
REFERENCE FCS LP 7-11-20, OBJ. 1.2 l 002000K100 ...(KA'S)
J QUESTION 2.03 (2.50)
With the Steam Generator Bl owdown System in a normal line up for 100% power 'l operation, answer the f ol l ow i n g :
- a. How will the bl owdown system respond to high radiation levels (above ;
alarm setpoints) due to a tube leak? Be specific. (0,5)
- b. Describe the release path that exists thr ough the blowdown system when it is lined up normal l y. Is release via this path monitored? -( 1. 0 )
- c. How should the bl owdown system be realigned to handle the contaminated I water. (1.0)
Y t
I h
t
. . . - . - - .c - - . . ,
,...-,.,.,,,_,._.__..m _ - . _ _ . . . _ _ . _ , . . . . - _ . , . . . _ _ . , _ _ _ _ . _ _ . . _ _ , . , _ , , . _ _ _ _ . _ _ _ , _ . . . _ . , _ , _ . . , _ _
2 ELadI_DESIEU_INCLUDlNB_SeEEIX_6MD_EMERBEUCY_SYSIEMS PAGE- 13' 4
ANSWER 2.03 (2.50)
- a. Bl owdown i sol a t i on v al ve s HCV-13G7A/B , 1388A/B shut.-(0.5)
]' b. Water entering the blowdown tank flashes to steam. Since the tank.is continuously vented to atmosphere, some of the steam and non condensable gases are released to the environment. (0.5)- This path is not monitored. (0.5) OR Water f rom bl owdown tank pumped by transfer pumps to RW header which is monitored.
- c. Isolate discharge to Raw Water (0,5) and line up to Waste D i sp osal System. (0.5)
P REFERENCE FCS LP 7-11-25 000037Ali3 ...(KA's) !
j QUESTION 2.04 (1.00)
- - Why do the auxiliary feedwater isaolation valves open when transferring aux i f eed con trol to AI-179?
l'
?
I '
ANSWER 2.04 (1.00) <
Operating the transfer switches momentarily interrupts DC power to the
- valve sol enoi ds. (0.5) This allows air to vent off the valve operator causing it to open. (0.5) -
, REFERENCE FCS LP 7-12-2 f 061000K601 ...<KA'S) i t
i QUESTION 2.05 (2.00) l Explain how to re-energize the RPS "M" coil following'a RPS trip initiated by:
l l a. Automatic RPS trip modules l
- b. Manual trip from AI-33 ,
i i,
P w -
.Y* v- --.,--=...,,-&- -- . - - . , _ _ , ______-%-._ .-
. .- - - . - ._ . _ ~ . -- . --
2a- ELAUI DESIGU.lGCLUDlGB 36EEIX BOD EGEEDEULX SXSIEUS PAGE -14 ANSWER 2.05 (2.00)
- a. Once the initiating trip has cleared (0.1), the operator must wait for KTD1 and KTD2 rel ays to time out. (0.4) The "M" coil is then re-energized by pressing the trip reset pushbutton. (0.5)
- 6. Manually close the trip breakers (0.5), press the trip reset pushbutton (0,5)
REFERENCE.
FCS LP 7-12-25 E0 1.16 001000K603 ...(KA'S)
QUESTI ON 2.06 (1.50)
For each of the following situations, indicate if the 480V breaker line ups are al l owe d . Assume all interlocks operate properly.
- a. T13-4C open 184C-open 183C closed BT-184C closed BT-183C closed
- b. 183A closed BT-183A closed BT-184A closed 184A closed
- c. T1848 open 1848 open BT-1838 closed i
L ANSWER 2.06 (1.50)
- a. yes
- b. no C. yes l (0.5 ea)
REFERENCE l FCS LP 7-13-3 EO 1.05 1 062000K407 ...(KA'S) i I
l
(+*++* CATEGORY 02 CONTINUED CU NEXT PAGE **+++)
!_.-_.__..___.~__,___.,.._____ _ _ _ _ _ _ _ _ _ . _ . . _ _ _ _ , _ _ _ _ _ _ _ _ _ _
'2e__ELLUI_DESIGu_1MCLUDING_SeEEIY_AND_EMERGEUC1_SYSIEUS PAGE 15 OUESTION 2.07 (3.00)
- a. State the power suppl i es avai l abl e to Safety Related 120VAC Instrument Power Inverter units and indicate the normal source. (0.75)'
- b. What are two methods by which the units can be switched to the other (al ternate) power supply? (0.5)
- c. HOW MANY and WHAT TYPE (Safety /Non-Safety) of 120VAC Instrument Power Inver ters would be di rec tl y impacted by the loss of a single 125VDC bus? (0.75)
- d. Besides the attributes discussed in a. and b. above, what other design feature adds reliability to the power supply for the Safety related 120VAC Instrument busses? (1.0)
ANSWER 2,07 (3.00)
- b. There are manual and static (automatic) transfer switches. (0.75)
- c. Two safety rel ated (0.5) and on non-safety related. (0.25)
- d. Each inverter output is arranged so it can supply one bus besides the bus to which it is dedicated. (0.75)
REFERENCE FCS LP 7-13-4 E0 1. 2, 1.3 062000K410 ...(KA'S)
QUESTION 2.08 (;.00)
Why is it important to reset PPLS and CPHS lockout relays BEFORE resetting SIAS lockout relays?
ANSWER 2.08 (1.00)
The SIAS Icckout relay coil will ' burn out" if it is reset before clearing the initiating signals (PPLS, CPHS). (1.0)
REFERENCE FCS LP 7-12-5 EO 3.04 013000K401 ...(KA'S)
(***++ CATEGORY 02 CONTINUED ON NEXT PAGE +++++)
. , , . .. - - .~ . . . _ . _ - . - _ - . .
2.__ELaNI_ DES 1CtLINCLUDluG_SaEEIY_aND_EtdERGEUCY_SYSIEt1S . PAGE 16 OUESTION '2.09 ~(2.00)
Briefly describe the design of the Reactor Vessel Level Indicating system and HOW it generates a vessel-level signal.
ANSWER '2.09 (2.00)
The RVLI system uses strings of paired thermocouples spaced at prescribed i n terval s and arranged vertically f r om the vessel head to the top of the active fuel. (1.0)
One'of the thermocouples in each pair is heated electrically while the other is not. As long as there is water surrounding a thermocouple pair
-the temperature difference between the two i s smal l . However, when the ;
water level drops below a thermocouple pair the temperature difference ,
r i ses. si gni f i can tly due to~the lower heat' transfer capability of steam.
This differential. temperature variation is used to establish vessel level.
(1.0)
REFER &4CE ,
FCS LP 7-12-23 EO 1.04 002000K402 ...(KA'S)
OUESTION 2.10 (2.00)
- a. What provides backup cocling capability to bring the plant to a safe shutdown condition if Component Cool ing Water is not operable? (1.0)
- b. What is the potential public hazard associated with the use of this ,
CCW backup system even if there is no loss of RCS integrity? (1.0)
[
ANSWER 2.10 (2.00)
- a. Raw Water (1.0) (supplied to various CCW cooled components individually)
- b. Failure to properl> i sol ate the CCW system prior to initiating RW would release toxic chemi cal s into the river. (1.0)
REFERENCE FCS LP 7-11-6 EO 5.3 000026A103- ...(MA'S)
(*++** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
4 t-r* ,w- ..-gy cr%.y-i%.+y e-%9m--,-re-m -m-- -,-e----- =cy-,--,-,i-vv,-~m- -----.-mw, r-, -+-w-,-- w -
_ . . _ . . _ _ . . _ . . - - . _ . _ _ . . _ . . _ . . _ _ . ~ ~ . . _ _ . _ . _ _ . . _ . _ _ _ . . _ . - - _ _ . . . _ . _ _ .
- i. J8 * . .
g 2.__oLANI_QESIGu_1NCLUDING_SeEEIX_AND_EMEEGEUCY_SYSIEMS PAGE 17-
- j. <
i.
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i . .
QUESTION 2.11 (1.00) !
f.
, How is the operation of the RM61 affected by placing the normal / inversion !
! switch In INVERSION? When is this position used?
I l
l ANSWER 2.11 (1.00)
I
- Placing the switch in INVERSION increases both the alert and alarm i setpoints by a fixed amount. (0.5) It is used when a radon inversion
[. (temperature inversion) has been confirmed since the radon wi>1 add to the '
j background level increcsing the possibility of erroneous alarms'and I component activation. (0.5) 7 REFERENCE FCS LP 7-12-3 EO 6.2 I
1 073000K101 ...<KA'S)
I t I QUESTION 2.12 (2.00)
I
- a. Give two conditions which will cause automatic operation of the ,
j electric fire pump. (1.0).
t- :
j '
i b. Give two conditions which will cause automatic operation of the diesel f
!, fire pump. (1.0) i-I i
ANSWER 2.12 (2.00)
- a. 1. Header pressure dropping to 109 psi. (0,5) ,
- 2. Whenever any transformer spray temperature bulb call s f or deluge .
valve to open. (0.5) i
- b. 1. Header pressure dropping to 79 psi. (0.5)
- 2. When the electric pump fails to restore pressere (or fails to a start) within 10 seconds of receipt of an auto start signal, t' (0.5) i REFERENCE FCS LP 7-11-12 f
) 086000A202 0860004301 ...(KA'S) i i
l I
) (****+ CATEGORY 02 CONTINUED ON NEXT PAGE +****) :
r I
t 5
i
2o__ELeUI_DESIGU_ldCLUDIUG_SeEEIY_eUD_EUEEGEUCY_SYSIEUS PAGE 18 CUESTION 2.13 (1.50)
List in order the three (3) actuations that are generated by the main condenser pressure instrument as vacuum decreases from 29" to 15" Hg at '
100'/ power. Setpoints are not required.
1 ANSWER 2.13 (1.50)
- 1. low vacuum alarm
- 2. turbine trip
- 3. steam dump / bypass inhibit
- 4. standby condenser evacuation pump starts (any three at 0.5 ea)
REFERENCE FCS LP 7-11-4 000051A202 ...(KA's)
, OUESTION 2.14 ( .50) i
! TRUE/ FALSE?
i l A reactor trip will occur when only one et breakers 3451-4 or 3451-5 is tripped.
I ANSWER 2.14 ( .50)
FALSE (0.5)
- 0000075.203 ...(KA's) l <..... Euo os CAree0Rt 02 . ...>
l i
1 l
L_ ___ ____ _ _ _ __ - - i
3._E1USIEduEUIS_auD_CCUIROLS PAGE 19 QUESTION 3.01 (2,50)
- a. What conditions will generate an.AFAS? (1.0)
- b. After. automatic initiation, what generator level will the auxiliary feedwater system maintain without operator intervention? (0.5)
- c. In order for the operator to control SG 1 evel s f rom CB-10 using aux feed after AFAS initiation, what is the required position of the switches for HCus 1107A/1108A and HCVs 11078/1108B on AI-66A/B? (1.0)
ANSWER 3.01 (2.50)
Steam press < 500 psi: (32% WR level to SG w/75 psi higher than other (0.5)
au t oma t i c al l y reset causing HCVs 1107A/B and 1108A/B to shut. AFAS :
re-initiates at 32% WR.) (0.5) .
C. 1107/SA OPEN (0.5) 1107/88 CLOSE (0.51 REFERENCE FCS LP 7-11-1 E0 1.2 1.4 013000K107 ...(KA'S)
I QUESTION 3.02 (2.00)
- a. What i s required for automatic CSAS initiation? (0.5)
- b. TRUE/ FALSE? CSAS initiation sends an open signal to the isolation valves but does not affect the spray pump status. (0.5)
- c. HOW and WHY does initiation of RAS after CSAS affect termination of containment spray? (1.0)
[
v E
- . - _ - --..__.__.._-,..__.____,_,_.,.__._.,.__m_-______,_,__,__ _ , _ _ _ _ _ , . , _ . _ . . _ _
~
3.__LUSIEUMEUIS_AUD_CONIROLS PAGE 20 t
s
{ ANSWER 3.02 (2.00) t ,
l a. PPLS AND CPHS . (0.5)
{ b.. true (0.5)
- c. RAS causes the spray pump recirc valves to shut. (0.5) Containment spray cannot be stopped by simply shutting the i sol at i on valves but requires reset of the other engineered safeguards signals (PPLS AND
! CPHS) to al l ow securing the spray cumps. (0,5)
~
b
! 026000A101 026000A103 ...(KA'S)
\
t i
OUESTION 3.03 (1.50) i
! A pressure detector in Steam Generator pressure channel 902 fails low.
! WHICH RPS trip functions are affected by this faiiure and HOW are they l l affected? !
- l 1 1 e
i l ANSWER 3.03 (1.50) ;
1 The trip functions for S/G. low pressure (0.5) and ASGT (0.5) are affected. l A si ngl e channel S/G low pressure trip and a single channel ASGT trip will i be indicated. (0,5) l REFERENCE FCS LP 7-12-25 EO 1.3 ;
012000A205 ...(KA'S) l QUEST I ON 3.04 '2.00) f
- a. Describe how the follcwing operate to manipulate the lineup or change "
functions of the Reactor Protective System. Give setpoints. (1.25)
- 1. Wide Range Log Channel [
- 2. Power Range Safety Channel i i
- b. Describe how the Nuclear Instrument System (NIS) functions to alert f the operater of a dropped CEA, (0.75) !
I i
l h
,--. ~ m _. . . . -. _ _ _ - - - -- - - - - - - - --
_. . _ _ . - . -- - _ _ - - =-. - . - - . . ~ . _. - . . . .. .
. '3._21NSIEDMEUIS_6UD_fDUIEDLS PAGE 21 l l
l i
/t45WER 3.04 (2.00)
- a. 't . Removes zero power bypass > 10 EE-4% power I Disable rate of change output < 10 EE-4% power.
I
- 2. Disable rate of change output > 15% p owe r .
Enable APD trip at 15% power.
Enable loss of load trip at 15% power.
(The reverse of each is acceptable. 0.25 ea) i
- b. Power range safety channel present output i s c omp ar e d to a time delayed previous output to develop and output a "CEA Dropped" signal.
OR Power decreases by more than 8% in < 8 seconds.
(Accept other viable NIS indications for no credit.) (0.75) !
i l REFERENCE !
FCS LP 7-12-25, 7-12-19 E0 1.6 i
015000K101 ...(KA'S) i
[ OUESTION 3.05 (1.50) h i
! At 100X power, Feedwater Regulating System in automatic, a high downcomer i l 1evel occurs on the 2A S/G. Assume no operator action. >
- a. What actions will occur with the FRS for S/G 2A while the high !
i downcomer level signal is input? (1.0) ;
- b. What will occur with the same system when the signal clears? (0,5) ;
l l I ANSWER 3.05 (1.50) t
- a. Controller shifts to manual (signal locks in) (0.5) and 'he FRV close. ;
(0,5) t
- b. "Lock in" manual signal is input when the open s i gn al clears causing the valve to open to the pre-high level condition position. (0,5) l REFERENCE j FCS LP 7-12-15 EO 2.3 !
059000K104 ...(KA's) l
(+++++ CATEGORY 03 CONTINUED ON NEXT PAGE **+++)
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- . . . - . . - . . ... ._ - - .. ._~ - - ... _.__ ~ _ _ .. . - . . . . ~ . ~ . - . -. -
t.'
3.__lMSIRudENIS_eND_CONIROLS PAGE 22 L
, 1 i
. l 1
OUESTION 3.06 (2.50)
In going from source range to full power, the WR NIS employs two swaps ,
t (changes in its operation) to provide continuous power indication. i t
i
! a. Describe the nature ofLeach swap and the power level at which it t
( occurs. (2.0) l i
f b. What will be the indication 1seen by the operator as each swap occurs?
-( 0 . 5' i
i ANSWER 3.06 (2.50)
{ a. At about 1000 CPS /4x10E-5% (0,5), the NIS goes from CPS to % power
}
(0.25) and shifts from 2 detecto's to 1 detector input (0.25).
At about 10 E-2% p owe r (0.5), the NIS shifts from a log count rate i l circuit to a campbelling circuit (0.5). ,
t
- b. For the first swap, the operator will see the CPS light go off and the
! % power lighi come on at CB-4. (The_ recorder on CB-4 will drop 1-2 t
I . decades when the swap occurs.) (0.25) l- For the second swap, the operator cannot readil y identify that the l
! swap has taken place. (0.25) g 4
i REFERENCE l FCS LP 7-12-18 EO 1.7
, 015000A103 ...(K4'S) ,
l t i QUESTION 3.07 (1.50) i i
- a. Excluding the test switch, what are two (2) conditions that will cause
! an emergency diesel generator to automatically start and accelerate to l I full speed? (1.0)
I i
- b. How are the auto start capabilities of an EDG affected by placing the l f 183 Master Emergency Switch in the emergency mode at AI-133 A/B? (0.5) i i
i ANSWER 3.07 (1.50) ;
)
- a. OPLS (0.5), Bus undervoltage (27 reisy) (0.5) l
. b. The auto starts are b1ncked. (0.5) i l
I l :
i
. ~ . - - . - - . . - . . - . . . . -. . . . . . ~ . . - - - . - - . .
3.__luSIRt fumrs_ gun _fDNIRDLS PAGE ~23 REFERENCE -
. 064000A301 ...(KA'S) i QUESTI&4 3.08 (3.00)
List four (4) conditions which automatically reduce letdown f l ow . For each condition, give'the component that actuates to reduce ficw. ;
ANSWER 3.08 (3.00) f (ANY 4)
CONDITION COMPONENT
- 1. High temp (470 deg.F) Shuts TCV 202 (letdown i sol ation valve) from' TIC 202 i 2. Pzr lui control system Modu l a t e s LCV-101 -1 and 2 to restore program level (letdown flow control viv) l t'
l 3. High letdown f l ow (155 gpm) Shuts HCV 204, letdown containment from FIC 212 isolation.
- . t
- 4. Letdown backpress control Modulates PCV 210 to maintain back :
l PIC 210 pressure at setpoint ;
^
- 6. Hi intermediate press.<550) Shuts HCV 204
- 7. Hi l e tdown press (190 psi g) Shuts HCV 204
}
> (0.5 for each condition, 0.25 for each component)
{
REFERENCE 7
'FCS LP17-11-2 E0 1.2 i 004000K101 004000K405 ...(KA'S) ,
[
OUESTION 3.09 (2.00)
What conditions will cause one or more Raw Water Pumps to start ,
automatically?
l t
o h
l
..-..-.-.-l
Jo__lUSIELUEUIS_6UD_EDUIEDLS PAGE 24 ANSWER 3.09 ( '2 . 0 0 )
- 1. If the running pump trips, the pumps on the opposite bus will start in standby. (1.0)
- 2. All four pumps will start when sequencing is actuated by PPLS, CPHS, or SIAS. (1.0) (SEQUENCING must be a part of the answer for fui; credit.)
REFERENCE FCS LP 7-11-19 E0 1.5 076000K402 ...<KA'S)
QUESTION 3.10 (2.00)
- a. What provides indication of RCS level / volume in the control room when the system i s dr a i ne d be l ow the pressurizer level taps? (1.0)
- b. Aside from indication, what are two specific uses of the signals gene ated by the low range pressurizer pressure detector? (1.0)
ANSWER 3.10 (2.00)
- a. Control room indication is provided by differential pressure transmitter (LI 197) betaeen a l ow pressurizer tap and a hot leg.
- b. 1. Shutdown cooling pressur,. interlock. (0.5)
- 2. Low temperature overpressure protection. (0.5)
REFERENCE FCS LP 7-11-23, 7-11-22, 7-11-20, SD I-4 002000K410 ...(KA'S)
QUESTION 3.11 (2.50)
- a. What three (3) conditions can cause a Rod Block?
- b. How can the operator reposition CEDMs as necessary to clear a Rod Bl ock ?
- c. What two (2) conditions / signals can cause a Rod Withdrawal Prchibit'
i- . .
.3o__lMSIEUMENIS_AND_CDUIROLS PAGE 25 i
l ANSWER 3.11 (2.50) 'l t
- a. PDIL, Deviation, or Out of' Sequence / Overlap. (0.25 ea) !
l
- .b. By controlling the affected CEDMs in maaual individual while holding the Rod Block Bypass Switch in "bypass". (0.75)
- c. Hi power pre-trip (0,5) ,
Hi SUR pre-trip (0.5)
REFERENCE FCS LP 7-12-26 E0 1.3 001050K401 ...(KA'S)
- GUESTION 3.12 (2.00) 1
- The NR NIS power level provides indication at the RPS and CB-4 plus
- actuation of the Level 1 and rod trip bistable on the drawer front. List .
four (4) other uses of the power signal from the NR NIS. i a
i l ANSWER 3.12 (2.00)
(ANY 4 AT 0.5 EA)
- 1. O power auctioneer
- 2. APD calculator
- 3. Comparator averager +
l 4. Deviation c omp ar i son and alarm I l S. Azimuthal tilt (Tq) i
! 6. Reactor Reg l 3
- 7. V0PT (HI PWR) i
. 8. Pl an t computer ;
- 9. TM/LP calculator l REFERENCE !
- . FCS 007-012-03-X-19-1.04-0001 -
l 015000K100 ...(KA'S) !
i i
! t i [
l l
l l t
i l i'
- (***** END OF CATEGORY 03 ****+)
4 f
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I d._;EROCEDURES_=_UDEMeL,_eBUDEdaL+_EUERGEUCY_auD PAGE 26 EaD10 LOGICAL _CQUIEOL QUESTION 4.01 (1.50)
- a. What are the normal OPPD permissible accumulated dose (PAD) limits as set by the Radiation Protection Manual? (1.0)
- 1. Daily
- 2. Weekly
- 3. Quarterly
- 4. Yearly
- b. By what factor are the daily and weekly PADS increased during major maintenance or refueling projects? (0.5)
ANSWER 4.01 (1.50)
- a. 1. 100 mr
- 2. 300 mr
- 3. 1250 mr
- 4. 4500 mr (0.25 ea)
- 6. 3 (three) (0,5)
REFERENCE FCS RPM VII-2 LP 7-9-3 EO 1.3 194001K103 ...(KA'S) 1 OUESTION 4.02 (1.50)
A Very High Radiation field exists in an area which cannot be controlled by locked access. What three (3) al ternate measures are required by Technical Specification section 5 to control access?
ANSWER 4.02 (1.50)
The area shall bei
- 1. roped off such that an individual ut the rope boundary is exposed to
(= 1000 mrem /hr.
- 2. conspicuously posted.
- 3. a flashing light shall be activated as a warning device.
(0.5 ea)
'.4 _ ERDEEDURES_ _dDEldGL 4_eENDRMAL . FMERDENCY eND PAGE. 27 RADlDLDBlCAL CDNIRDL REFERENCE FCS TS 5.11.2, LP 7-9-3 EO 2.0 194001K10s ...(KA's)
QUESTION- 4.03 (1.00)
The 161XV system becomes inoperable while the plant is at 100% power and is expected to remain so for at least 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. What actions are directed by Techn i c al Specifications to be performed immediately?
ANSWER 4.03 (1.00)
- 1. Verify remaining power source
- 2. Notify NRC (by telephone or telegraph)
(0.5 each)
REFERENCE FCS LP 7-13-1 EO 1.8 ,
062000SG3 ...(KA'S)
QUESTION 4.04 (2.00)
OI-VA-2 provides guidance on how to isolate and identify a leak in the Aux building whict' caused a Stack High Radiation Alarm. Briefly describe these actions.
! ANSWER 4.04 (2.00)
- 1. Secure supply and exhaust fans
- 2. isolate all rooms j 3. start DNE exhaust fan
- 4. open dampers one room at a time until the leak is found (area rad monitors may also aid in leak isolation)
(0.5 ea)
REFERENCE FCS LP 1-14-1 EO 1.6, OI-VA-2 .
I 073000SG8 ...(KA'S) l
(+++** CATEGORY 04 CONTINUED ON NEXT PAGE ++**+)
l l l
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a __'EROCEDURES_ _UDEt%L,_eBUQEtML,_EMEEEENCY_eUQ PAGE 28 red 10LOGLCeL_CONIROL-i i
QUESTI ON 4.05 ( .50)
Choose the letter which corresponds to the best answer. ,
The purpose of Operating Instructions (Ols) is to provide:
- a. a detailed procedure for the operation of plant systems and equipment. '
- b. a detailed procedure for the operation of specific equipment only.
s c. general instructions for the operation of particular systems.
- d. general instructions for integrated operation of the plant.
ANSWER 4.05 ( .50) !
E
, A (0,5) i REFERENCE
! FCS S.O. O-17, LP 7-15-7 EO 1.1 !
L QUESTION 4.06 (1.50) [
AOP-11 directs you to TRIP THE REACTOR if CCW flow cannot be restored !
I within five minutes or a high temperature alarm is received on any of three ,
- parameters. List the three (3) parameters. [
< r l
j ANSWER 4.06 (1.50) !
- 3. RCP soal bleed off temperatures !
(0.5 ea) {
! REFERENCE FCS LP 7-17-11 E0 1.2, AOP-11 000026K303 ...(KA'S) ,
- e 1
4 i QUESTION 4.07 (1.00)
I l Briefly explain the limitation associated with establishing charging and j letdown as an al ternative means of residual heat removal following a loss ;
of shutdown cooling per AOP-19.
' I l !
i
- . , . _- . _ _ _ _ - - _ _ . , . _ _ _ . . - . - . _ . - _ .- J
. . . . _ ~ =~ . . . - - _. .. . -. . ..
a.__EROCEDURES_ _UDEdeL,_6EUGEMAL,_EMEEGENCZ_aMD PAGE 29 ReDlOLOGICAL_CQUIEDL ANSWER 4.07 (1.00)
DJe to the low flow rate available f r om the charging pumps (0.5), this
-method will only be effective many days after shutdown when decay heat production has decreased. (0.5) i REFERENCE FCS LP 7-17-19 E0 1.0, AOP-19 0050009310 ...<KA'S) t t
I r
i QUESTION 4.08 (2.00) i According to AOP-22, "Reactor Cool an t Leak," explain four (4) way to verify that the code safeties or PORVs are not responsible for the loss of RCS inventory.
I j ANSWER 4.08 (2.00)
(ANY 4 AT 0.5 EA)
- 1. Normal quench tank temp
- 2. Normal quench tank press i 3. Normal quench tank level
- 4. No relief header temp alarms
- 5. No relief header acoustic flow alarms REFERENCE FCS LP 7-17-22 E0 1.0, AOP-22 002000K405 ...(KA'S)
I L
i OUESTION 4.09 (1.00)
What is the operator's first action to attempt to terminate an uncontrolled i CEA witndrawal?
i i
ANSWER 4.07 (1.00) i 4
Place the CEA drive control selector switch to "0FF" position. (1.0)
L REFERENCE
, FCS LP 7-17-27 E0 1.1, AOP-27 000001k301 ...(MA'S)
I (*++** CATEGORY 04 CONTINUED ON NEXT PAGE +++++)
n
_ _ . . _ . _ _ . - . . =__.s- _ _ _ - . _ . _ _ _ _ . _ . . . ~ _ . _ _ . - - _ _ . _ . _ . _ _ . _ . _ _ _ _ . . . _ - . _ .
30
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d.__EEOCEDUEES_=_ NORMAL +_eBUQEdeL+_EUEEGEUCY_eUD PAGE
- . R&DIOLOGICaL_CQUIROL I
i QUESTION 4.10 (2.00) .
Per EOP-01, how is.the RCS heat removal. safety function verified (two answers required)?
f AN SWER . 4.10 (2.00)
- 1. Verify at least on S/G has level
l a. 35-85% NR with feedwater available (0.5)
- b. Being restored by adequate feed flow. (0.5)
! 2. Verify steam and bypass valve functioning to control:
- a. RCS to 525-535 Deg.F (0.5)
- b. S/G pressure 850-925 psia (0.5) ,
REFERENCE I FCS LP 7-18-1 EO 1.3, E0P-01
- 000007K301 002000K111 ...(KA'S) ,
i !
1 ;
QUESTION 4.11 (1.50)
If reactivity control is not established per E0P-1, "Reactor Trip," what [
two (2) contingency actions must be taken by the operator?
j l I i I
ANSWER 4.11 (1.50) i-
- 1. Open CEDM clutch power supply breakers on panel AI-57 (0.75)
Emergency borate (0.75) i
!. 2.
I REFERENCE
, FCS LP 7-13-1 EO 1.4, EOP-1 t
- 000024K302 ...(KA'S) f 4
i i
, QUESTION 4.12 (2.00)
During Reactor Trip Recovery, the operator is cautioned not to overreact to
' low S/G 1evel and add feedwater sl owl y to avoid problems. What are four
. (4) of the potential problems to be avoided during recovery ct S/G 1evel? !
! t i
l
?
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fL__EsOCEDURES - NQEtdAL , _aEUORtSL,_EtdERGEUCY._.AUD PAGE 31 RAD 10 LOG 1CeL_CQUIEQL ANSWER 4.12 (2.00)
~1 . Excessive cooldown
- 2. Excessive transient in pressurizer _ level and pressure 3 ~. Overfi11ing S/Gs
- 4. Pocnibility of. feed ring damage / water hammer if feeding via feed ring and S/G level dtups below feed ring.
(0.5 ea)
REFERENCE-FCS LP 7-18-1, E0 1.4, EOP-1 000007K106 000007K301 ...(KA'S)
QUESTION 4.13 (1.00)
Explain why the oper ator is directed to cool down be l ow 510 deg,F prior to i sol at i ng the affected S/G when f ol l owi ng EOP-04, SGTR.
ANSWER 4.13 (1.00)
Reducing RCS Th to <= $10 deg F will maintain adequate RCP NPSH and RCS subcool ing when RCS pressure is reduced below S/G safety valve setpoint.
(1.0)
REFERENCE FCS LP 7-18-4 E0 1.12, E0P-04 000038K306 ...(KA'S)
(+++** CATEGORY 04 CONTINUED ON NEXT PAGE ++***)
L._._ERO CEDU RES _ = _U G EMa L +_eBU Q EMA L +_EME R G ENCY _eN D PAGE 32-EaQ10 LOGICAL _CQUIROL 6
QUESTION 4.14 (2.00)
For each item, arswer TRUE or FALSE concerning OI-DG-2, "Normal Operation, Diesel Generator No. 2":
- a. Diesel-Generators may be tested concurren tly if one diesel is maintained running for one hour after shutdown of the other.
- b. The Diesel Generator can be loaded when jacket water temperature reache s 100 deg.F.
- c. Pl ac i ng the master emergency switch (183/mes) at AI-183A/B out of the normal position will transf er control of the diesel to the local control panel.
- d. A diesel generator. should not be star ted or loaded in anticipation of a loss of offsite power.
ANSWER 4.14 (2.00)
- a. false
- b. f al se
- c. true
064000K102 064000K303 064000K401 ...(KA'S)
QUESTION 4.15 (2.50)
The f ol l ow i n g r e l a t e to OP-3, "Plant Startup from Hot Standby to Minimum Load."
A. List the three (3) reactor trips which are monitored as power is increased above 15% power and state what the operator i s required to verify,
t
- ' = .
5L__EB O C EDUEES_=_tJOEMAL,_.eEUQEt:!a L,_ Et:!E R G ENCY_.eUD PAGE 33^ ,
EaD10 LOGICAL _CQUIEOL '
ANSWER 4,15 (2.50)
) a. 1. Lsss of Load - bypass removed /tr?p enabled
- 2. APD - bypass removed / trip enabled !
- 3. High rate of power change - trip bypassed / disabled i (0.5 ea)
- b. To minimize thermal and reactivity transients while at low pourr. (1.0 REFERENCE i FCS LP 7-51-2 EO 1.1, 1.3, OP-3 002000K507 002000K511 072000A301 . . < !(A's)
W GUESTION 4.16 (2.00) :
1 List four (4) of the actions an operator must take if the reactor becomes cr i t i cal be l ow the ECP while performing OP-7.
ANSWER 4.16 (2.00) !
. (ANY 4 AT 0,5 EA)
- 1. Make the reactor subtritical (insert CEAs)
- 2. Place all 4 Extended Range Cutout switches in "off" i
- 3. Notify the Reactor Engineer (or designee)
- 4. Verify bcron cc7 centration, time and Tavg
- 5. Recalculate the ECP l 6. With Reactor Engineer's permission, resta't the reactor, otb)rwise go ;
l to hot shutdown and notify Manager-FCS ( or des!gnee)
REFERENCE l FCS LP 7-51-2, E0 1.3, CP-7 !
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t Omaha Public Power District 1623 Harney Omaha. Nebraska 68102 2247 ,
40 2536-4000 ' 1 7)
!ii February 17, 1988 LIC-88-117 I Ii 62 XO
.I D i HAR24tg l.JW Mr. Robert Martin Regioaal Administrator Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive Suite 1000 Arlington, TX 76011
Reference:
Docket No. 50-285 i L
I
Dear Mr. Martin:
The Fort Calhoun Station staff has completed review of the Reactor / Senior >
Reactor Operator license examinations administered on February 9, 1988.
It is requested that the enclosed consents be considered in grading the examinations.
Please feel free to contact Jay Fluehr at (402) 533-6730 if you have any ,
questions.
Sincerely,
'/ N,/
R. L. Andrews Division Manager Nuclear Production RLA ket Enclosires c: LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Ave., N.W.
Washi>gton, DC 20036 Jose A. Calva, Project Director Anthony Bournia, NRC Project Mgr.
John Pellet, Operator Licensing F. H. Harrell, NRC Resident Inspector ,
4s se _ _ ___ _ -- - - _ -__.m .~_m_ -
_=_-_ __ _ _ . _ . _ _ _ _ _ _ _
i
t FORT CALHOUN STATION REACTOR / SENIOR REACTOR OPERATOR LICENSE EXAMINATION COMMENTS QUESTION. 1.04/5.04 The answer does not address automatic actions which would occur because of the dropped rod. No credit should be deducted for de-scribing the following plant response:
- The decreasing temperature will cause a corresponding de-crease in RCS pressure. If the pressure decreases to 1700 psia, the reactor will trip due to TM/LP.
- When steam pressure decreases to 500 psia, the reactor will trip due to low S/G pressure and SGIS will close the MSIVs, thereby terminating the cooldown.
Reference:
FCS LP 7-12-25, Revision 1, page 93 OUESTION 1.05 :
Since the question asks for general actions, we believe that "in-crease temperature" should be considered an acceptabla answer.
QUESTION 1.10/5.10 The first sentence of the answer stands alone and res'onds to t'do question of why a void may form in the head. The next two sentences describe h2w a void forms. It is recommended full credit be given for the concept in the f.irst sentence.
l 1
l l
OUESTION 2.03/6.02 j The correct answer to Part a. should be: l 1
A high alarm on either Blowdown Radiation Monitor isolates blow-down from both steam generators (RM-054A shuts HCV-1387A and HCV-1388A; RM-034B shuts HCV-1387B and MCV-1388B) . Valve num-bers and Radiation Monitor numbers should not be required.
l The complete answer to Part b. should include another release <
flowpath:
Water from the Blowdown Tank is pumped, by Transfer Pumps, to the Raw Water discharge header. The Raw Water discharge header is monitored.
Reference:
FCS LP 7-11-25, Revision 1, pages 39, 51 & 52 FCS LP 7-11-19, Revision 1, pages 58-59 OUESTION 2.07 It is suggested the point value for part d. is too high to be consistent with its importance to safe plant operation. It is suggested that the points be redistributed to make each part of the question worth .75.
QUESTION 2.08/6.03 Since SIAS will not reset until PPLS and CPHS are reset, a bet-ter way to ask this question on future tests would be:
What is the possible adverse effect of attempting to roset SIAS prior to resetting PPLS and CPHS?
OUESTION 2.11/6.06 The answer states that the normal / inversion switch is used when a radon inversion has been confirmed. Temperature inversion is the terminology used in Fort Calhoun Station Operating Instruc- ,
tions.
Reference:
OI-RM-1 I.E.
2
-QUESTION 2.13 The correct answer should be any three of the following (in cor-rect order) ':
- 1. Standby Condenser Evacuation Pump starts
- 2. Low vacuum alarm
- 3. Turbine trip
- 4. Steam dump / bypass inhibit .
Reference:
FCS LP 7-11-4, Revision A, pages 51 and 52 FCS LP 7-11-5, Revision 0, II, C. 2. d. (10)
(enclosed)
OUESTION 3.01/6.08 The answer for part a. is correct except that loss of power se-quencer S1 and S2 do not cause an AFAS and should be deleted from the answer. This le a lesson plan error for which a re-vision has been initiated.
Reference:
T.S. Table 2-3, No. 5 OUESTION 3.03 There is a single S/G low pressure trip unit (shared by both S/G's) for each RPS channel. There is also a single ASGT trip unit for each channel. Ther-* ore, the correct answer should be: The trip functions for S/G low pressure (0.5) and ASGT (0.5) are affected. A single channel S/G low pressure trip and a single channel ASGT trip will be indicated.(0.5)
Raference: FCS LP 7-12-25, Revision 1, pages 67, 78 & 79 OUESTION 3.04 A power decrease exceeding 8% in 8 seconds should be considered another acceptable answer to Part b.
Referenca: OI-NI-1, IV.D.6 3
1
OUESTION 3.05/6.09 The construction of the question does not elicit the detailed response shown in the answer key. The acceptable answer should include:
- a. The controller shifts to manual and the FRV closes.
- b. The valve opens to the pre-l'igh level position (and is in manual control).
Reference:
OI-FW-3, page 3, D.4 OUESTION.3.06/6.1Q Another correct response for part b. of the answer is:
The strip chart recorder indication on CB-4 shows a drop of 1-2 decades when the swap from CPS to % power occurs.
Referance: FCS LP 7-12-18, Revision 1, page 55 QUESTION 3.08 Instrument numbers and reason for each action (e.g., to restore program level) are not elicited by the question and should not be required. Two other answers should be acceptable:
CONDITION COMPONENT High (intermediate) letdown pressure Shuts HCV-204 (550 psig)
High letdown pressure (190 psig) Shuts HCV-204
Reference:
FCS LP 7-11-2, Revision 1, pages 36 & 37 OUESTION 3.09 Since the question does not ask how many pumps or which pumps would start, full credit should be given for describing the conditions that cause automatic starts.
"PPLS, CPHS, or SIAS" should not be required for full credit in Part b. if the examinee describes that they are started by the sequencers.
Reference:
FCS LP 7-11-19, Revision 1, page 85 4
l
- O QUESTION 3.12 Additional correct answers should be:
- Variable Over Power (high power) trip
- Plant computer TM/LP calculator
Reference:
FCS LP 7-12-19, Revision 1, page 14 OUESTION 4.01/7.01 The yearly PAD is 4500 mr. This is a recent change to the Fort Calhoun Station Radiation Protection Manual.
Reference:
Radiation Protection Manual, VII-2-8, 2.15.1 (enclosed)
OUESTION 5.03 The question may tend to confuse the examinee, since it does not specify whether Peak Linear Heat Rate or Average Linear Heat Rate is referenced to. Two additional answers should be considered correct:
- 1. Ths linear heat rate increases in the part of the core that was shadowed by rods. l
- 2. The core average linear heat rate does not change since reac-tor power is kept constant.
Reference:
FCS RX Theory LPs 7-5-3, 7-5-4 & 7-5-5 OUESTION 8.04 Completion of the work should be considered an acceptable an-swer. Mention of Maintenance or Design Change Order should not be required.
Reference:
S.O. M-9 5
. o OUESTION 8.06 Another correct response to Part b. is:
When the Licensee declares one of the Emergency Classifications.
Reference:
S.O. R-11, para 2.1 and Appendix A, A.l.a.
QUESTION 8.07 In the answer to Part b, the Reactor Operator and Assistant RO may also be called the Licensed Senior Operator (LSO) and Li-censed Operator (LO).
Reference:
Shift Turnover Log, FC-95 (enclosed)
OUESTION 8.09 Prevent inadvertent actuation of HPSI which could overpressurize the RCS should be an acceptable answer to Part a. The discus-sion of brittle fracture should not be required since it is not included in the Technical Specification basis.
Reference:
T.S. 2.3 OUESTION 8.11 In Part c. of the answer, Main Steam Safety Valves is acceptable terminology at Fort Calhoun for S/G Atmospheric Reliet Valves.
Reference:
OI-MS-1, III.A.
QUESTION 8.12 In the snswers to Parts a. and b. the Assistant RO and the Li-cenaed Equipment Operator may be called Licensed Operators (LO),
Reference:
Shift Turnover Log, FC-95 (enclosed)
J i 6 4