ML20127G752
| ML20127G752 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 04/16/1985 |
| From: | Rogers T, Wilson B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20127G713 | List: |
| References | |
| 50-395-OL-85-01, 50-395-OL-85-1, NUDOCS 8505210135 | |
| Download: ML20127G752 (123) | |
Text
ENCLOSURE 1 EXAMINATION REPORT 395/0L-85-01 Facility Licensee: South Carolina Electric and Gas Company P. O. Box 764 Columbia, SC 29218 Facility Name: V. C. Summer Facility Docket No. 50-395 Written, oral, and simulator examinations were administered at the V. C. Semmer Nuclear Plant near Jenkinsville, auth Carolina.
Chief Examiner: u ., - 4, s s _c F//6/Ps Thomas Rogers Date Signed Approved by: '
GlAbd V//G/ff f Bruce A.' Wilson, Section Chief Da'te 'Si gned Summary:
Examinations on February 25-28, 1985 Examinations were given to four SRO candidates two of whom passed; five R0 candidates, two of whom passed; and one instructor certification, who failed.
8505210135 850419 PDR G
ADOCK 05000395 PDR k
REPORT DETAILS ,
- 1. . Facility Employees Contacted l
- Connelly, J., Deputy Director of-Operations and Maintenance !
- Heilman, J. F., Associate Manager of Nuclear Operations Training i
- Matlosz, T. L., Nuclear Operations Training Supervisor '
- Williams, M. B., Manager - Nuclear Operations Education and Training ;
- Attended Exit Meeting
- 2. Examiners Hemming, B.
-Jaggar, F.
Picker, R.
- Rogers, T.
- Chief Examiner
_3. Examination Review Meeting At the conclusion of the written examinations, the examiners met with Messrs. Terry Matlosz, Frank Blanchard, _ Tim Cox, and Russell Bender to review the Reactor Operator written examination and answer key. Following the Reactor Operator _ exam review, Messrs. Terry Matlosz, Frank Blanchard, Randy Ruff, and Russell Bender reviewed the Senior Reactor Operator written examination and answer key. The following comments were made by the facility reviewers:
- a. SRO Exam
- 1. Question 6.8 Facility Comment:
"As of April 19, 1984, the Automatic Reactor Trip Breaker actuation signals will.deenergize the U. V. coil and energize the shunt coil i associated with each reactor trip breaker. Answer "C" is incorrect and Answer "A" is a correct answer as the Automatic trip signal will also energize the shunt coil.
A.
References:
- 1. MRF 20208
- 2. GAI Drawing: IMS-43-003 IMS-43-002 IMS-43-189"
t i
2 NRC. Resolution:
The additional references provided indicate that the lesson plan does not reflect the change made on April 19, 1984. The answer key has been changed from "c" to "a" as the correct answer.
- 2. Question 6.20 Facility Comment:
"We request to delete this question based on the following. This question has no correct answer per the selections A-D. The answer given "A" (S/G "A" Press. 650 psig with S/Gs "B" and "C" at 700 psig) was not correct. Low pressure SI on 675 psig low main steam pressure is for 1/1 detectors on 2/3 S/Gs not 2/3 detectors on'1/3 S/Gs.
References:
Drawing 1080837" NRC Resolution:
The referenced drawing supports the facility contention that there is no correct answer provided. The question has been deleted from the examination.
'3. Question 6.27 Facility Comment:
"We request that this question be deleted on the following bases.
This question has no correct answer per the selections A-D. The answer given "A" (RHR Pump) is incorrect for this question per the following:
- 1. Drawing 04-4461-D-802-036
E0P-6.1; Page 6 of 16 Step 11 (80)
Conclusion:
You will find that D-802-036 shows an RHR pump start for both SI and/or blackout. Both E0Ps also support the fact that RHR pumps start for both SI and blackout."
NRC Resolution:
The additional references provided indicate that the lesson plan is incorrect and that no correct choice was provided. The question has been deleted from the examination.
3
'4. Question 8.18 -
Facility Comment:
"We. request that this question be deleted on the following bases.
This question' is' in conflict .with SAP-205 which doesn't require mode memorization as this question would have us do. This Tech.
Spec. item is a 30 day action requirement.
A.
Reference:
SAP-205 6.0 PROCEDURE 6.1 Removal from Service 6.1.1 The Shift Supervisor shall determine the following-upon discovery of an IN0PERABLE component or system or upon receipt of' a request to remove a component or system from service:
A. The function of the affected component or system. The system (s) affected. How AFFECTED may be a short statement or reference to a Tag Out Sheet.
B. The Technical Specification Operability requirements for the present plant mode.
C. Current status of the plant and the affected system or related systems requiring its operability. I D. Time limit for restoring. the component or system to service per the Technical Specifi-cation Action Statement.
This procedure doesn't require the Shift Supervisor to memorize A-D above, but only to determine A-D above based on his training and knowledge of the plant. By Station Policy Tech Spec- items in duration of < 1 Hr. should be memorized, due to this relatively short duration of time. Question 8.18 would have us memorize 30 day Tech Spec items."
NRC Resolution:
The question does not elicit any response for required operator actions upon discovering inoperable hydrogen- mitigation systems.
Furthermore, this question does not require memorization of modes and the applicable status of equipment if the candidates under-stand the bases for having combustible gas control systems. For
4 these reasons and because of the safety significance of these systems, the ' question- will remain as part of the examination as presented to the facility during the review.
- b. R0 Exam
'1. Question 2.21 Facility Comment:
~
" Answer "C" is incorrect for this question as per answer key given. Answer "D" is correct for this question as stated.
References:
- 1. GIA Drawing D-802-004" NRC Resolution:
The comment was not accepted as stated because the facility comment has transposed "c" and "d". The correct answer is "c" and the answer key has been changed to reflect the correct response as supported by the additional reference material provided.
- 2. Questions 3.07 and 3.08 Facility Comment:
"Both these Questions deal with COPS which is an obsolete system and does not exist. COPS has been replaced by RHR suction relief overpressure protection. These questions should be deleted.
References:
- 1. Technical Specification 3/4.4.34; Amendment No. 26 dated 9/24/84
- 2. MRF-20249
- 3. Drawings: a) E-210-208 b) E-210-207 c) S-212-082" NRC. Resolution:
The additional references provided indicate that the lesson plan has not been updated for the modifications made for cold plant pressure protection. Questions 3.07 and 3.08 have been deleted from the examination since they elicit knowledge of a system that no longer exists.
l I
1
_ ]
m-i 1
- 3. Question 3.14 Facility Comment:
"This question has parts A-E, but the answer key has parts A-F.
Due to the fact that the point value of the answers for A-F were
~ based on .5 points each, we request that ' Sections A-E be reassigned point values of .6 points each. This would keep the total point value for the question the same. This question,only has 5 parts and not 6 parts; therefore, needs to have point values reassigned to .6 each."
NRC Resolution:
Comment accepted. Due to a typographical error, the point values '
were redistributed for the expected responses.
- 4. Question 3.20 Facility Comment:
"We request that you delete the question based on the following:
Depending on the interpretation of the question, there are 3 correct choices for this question. Per Attachment IV-10.18, steam generator mass goes down from 0-30% power then ramps up as steam generator level ramps up from 30%-100% power.
- 1. Answer 1: If steam generator level ramps up from 30%-100%
power, then steam generator mass increases from 30%-100%
power.
- 2. Answer 2: Taking the other view that you are ramping down from 100% to 30% power, then steam generator mass decreases from 100% to 30%.
- 3. Answer 3: If you look at just 0-30% power, then steam generator level remains at 38% with mass ramping down to a low point at 30% power.
Conclusion:
The question is very general in what it wants.
The question asks how steam generator mass changes over the range of steam generator level ramp. It depends on whether you ramp up or down. Also, the only part that steam generator level is ramped is from 30%-100% power. From 0-30%
power, steam generator level is constant at 38% with steam generator mass ramping.
References:
- 1. Attachment IV-10.18
FJ g
6
- 2. IC-2 Steam Generator Water Level Control Figure IC2.2 (Steam Generator Program Level)"
NRC Resolution:
The facility request has not been accepted. There is only one correct answer for this question. The facility contention was based on only the first part of the question distractors. If the entire distractor is examined, the only correct answer is "d". .The question was directed at the programmed ramp of the level control system.
Answer "a": If viewed as a down ramp the water mass could decrease but the steam mass would also decrease, whereas an increase would cause water mass to increase and steam mass to increase.
Therefore, this answer is incorrect.
Answer "b": The water mass change is viewed as for "a" above but the steam mass _does not remain the same in the steam generator on either up or down maneuvers. Therefore, this answer is incorrect.
Answer "c": On an up power ramp water mass increases and so does steam mass, whereas on a down ramp, water mass decreases and steam mass also decreases. Therefore, since the water mass goes opposite of steam mass, this answer is incorrect.
Answer "d": As per the statement on answer "c", this answer is correct for an up power ramp because both parameters change in the same direction. Therefore, this is the correct answer.
No changes have been made to the examination or answer key.
- 4. NRC Post Grading Review Following the review of graded examinations in accordance with NUREG-1021, ES-108, Quality Assurance Program for Review of Graded Examinations, the following changes were made to the answer key and the effected examinations were regraded accordingly. <
-The following question was deleted from the R0 examination:
Question 4.08 Reason: The question was worded in such a way that one, two, or three of the four choices could correctly answer the question.
Since this possibility existed, the question did not elicit the required candidate knowledge of the subject matter.
No changes were made to the SRO examination.
u
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- 5. Exit Meeting l At the conclusion. of the site visit the examiners met with'_ representatives
'of . the ' plant . staff. to discuss the - results of the examination. Those individuals who clearly passed the operating examination were identified.
- There :was no generic weakness.noted during the oral examinations. Areas of 4
, below-normal performance were'.noted from the simulator. examinations for communications, event. diagnosis, and procedure' familiarity.
~
The cooperation given' to the examiners and .the effort sto ensure an atmosphere in-the control room conducive to oral examinations was also noted
-and appreciated.
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U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION J
FACILITY: _SydHER__________________
REACTOR TYPE _ EM R-M EC1__ __ ____________
2- D AT E A D MI N I S T E R E D t _R240ZLZ1________________
'- EXAMINER B.
_RI$EgR1 __________
4 APPLICANT ___ _
-INSIEUCIIGHS ID_AEELIC&dIl Uso separate paper for the answers. Write answers on one. side only.
Staple question sheet on top of.the answer sheets. Points for each qucstion are Indicated in parentheses af ter the question. The passing grade' requires.at least 704 in each category and a final grade of at lecst 80%. Examination papers will be picked up six (6) hours after tho examination starts.-
% OF I CATEGORY % OF APPLICANT'S CATEGORY l
__MALUE_ _IGIAL ___1COBE___ _M ALUE__ ______________C A IEGQE1_____________ l l
.22aQQ__ _ElaQQ ___________ ________ 1.- PRINCIPLES OF NUCLEAR POWER '
PLANT OPERATION, THERMODYNAMIC $s HEAT TRANSFER AND FLUID FLOW
.22aQQ__ _22aQQ ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY ~
AND EMERGENCY SYSTEMS
_12sQQ _ZiaQQ ___________ ________ 3. INSTRUMENTS AND CONTROLS
_AlaQQ__ _22409 ___________ ________ 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL IQQaQQ__ IQQaQQ ___________ ________ TOTALS FINAL GRADE _________________%
All work done on this examination is my own. I have neither givon nor received aid.
APPLICANT'S SIGN ATURE f
- 16__ESINClfLE1_DE_duCLEAE_EDMER_EL&HI_DEEEALIDHg PAGE 2 IMERBDDISABICit_dEAI_IE&ESEEE_ASQ_ELMIQ_ELQW QUESTION 1 01 (1.00)
' Approx sma tely how many hours do es it take for Xenon to reach 100%
equilibrium concentration af ter the r eactor is b rought to full power from a Xenon f ree condition? {
- a. 60-80 hours '
- b. 40-60 hours
- c. 20-40 hours M
- d.80-100 hours (1 0) -
QUESTION 1.02 (1.00) 9 If the reactor tripped after a 30 day run at 50% power, E0Ls what I would be the peak reactivity value of Xenon in PCMT
- c. ~ 1486 4
~
- b. 3950
' ~ ~
- c. 3700
! d. 1332 (1.0)
Q UES TION 1 03 (1.00)
Wha t happens to the worth (PCM) of the most reactive stuck rod as the core ages? -
- a. The worth ch an ge is not considered as the core ages.
- b. Remains the same c.- In cr e as es
- d. Decr e as es (1 0) l' a
=%
- La__ERINCIELES_QE HUCLEAE_EQtER_ELAHI_DEEE&IIDHs PAGE 3 ISEREQ21HABICSz_BEAI_IBASSEEE AHQ_ELUIQ ELQM Q UES TION 1.04 (1.00)
-As the core ages, the -r atio of PU239 atoms to U235 atoms incr eases.
This changing ratio causes: ,
u
- a. reactor period to decrease. j
- b. the. Void Coef fi ci ent becomes less negati ve. ,
- c. Moder ator Temper ature Coef f ic ient to become l ess n egative.
- d. the delayed neutr on f rac tion to increase. (1 0) -
?
- =
QUESTION 1.05 (1.00) Q s
The mar gi n to DNBt g
- a. Increases with decreasing f l o w r at e -
- b. Decreases wi th dec reasing reactor power i
- c. Decreases with decreasing pressurizer pressure i
- d. Increases with increasing Tavg (1.0)
Q UESTION 1 06 (1.00)
D ur ing a st e am leak from the main steam header to the atmospher e.
I-4 A throttl ing process is created. Which process below will occur?
j_
- a. Enthalpy o f the steam will de crea se . %
i
- b. Entropy of the steam will i nc rease. E a
- c. Speci fi c vo lume of the steam will decrease.1 l
- d. Steam temper atur e wi ll rema in the s ame. (1.0)
?
e b
l 4
+
1
- ls__ERINCIELIS_QE_HUCLEAR_EQtER_ELAHI_DEEE&ILOMt PAGE 4 IMEE50Q2H&51C34_ME&I_IRAMIEER_AHD_ELUIQ_ELQM i
I Q UESTION 1.07 (2.00)
Indicate on your answer sheet whether the following st at ements cro TRUE or FALSE.
- a. P um p runout is t he t er m used to desc ribe the condi tion of a cen tr i fu gal pump running with NO volume flow r at e. (0.5) '.
- b. If the speed of a centrif ugal pump is doubl ed, the fl ow r ate AND dis char ge pr essure will d o ubl e. (0.5) l
- c. If the speed of a positive displacement pump is doubleds the flow r ate will double. (0.5)
- d. For two centrif ugal pumps in SERIES, the comb ined delivery flow r ate is equal to the sum of the Individual pump flow r at es at the same pump speed. (0.5) l QUESTION 1.08 (1.00)
Ccnsider the f ollowing statements concerning un it ef fi ciency at a s te ad y state power level. Using only the paramenter change I ndiciated, choose the MOST CORRECT statement.
- a. Uni t e f fi ci ency decr eases if the current being d r awn by t he Reactor Coolant Pumps increases slightly due to a change in bus voltage.
l
- b. Uni t- ef ficiency increases if ABSOLUTE condenser pressure changes from 1 psia to 1.25 psia.
I c. Unit efficiency remains the same if total Steam Generator Blowdown flowrate is changed f rom 35-40 gpm.
- d. Unit efficiency increases if hotwell temper ature rises f rom 90-100 F. (1 0)
I
I 1.__ERINCIELE1_QE_HMCLEAE_EDEEE_fLABI_QEERAIIDHg PAGE 5 IMER50DINAfi1C3a_BEAI_IRAH1EER_ASD_ELUIQ_ELQM
! QUESTION 1.09 (1.50)
'Two identical reactors are started up using rods. Reactor A has a red speed of 50 steps per minute and Reactor B has a rod speed of 25 steps per minute. Assume a continuous rod withdrawal.
- a. Which reactor will go cr i tical -fi rst? (0 5) .
- b. Which' reactor will have the highest source range counts at cri ticali ty? (0.5)
.c. -How will the critical rod heights compare in the two reactors? (0.5)
QUESTION 1.10 (1.00) 4 Tho reactor has been started up on a new core and has achieved 1005 pceer with equilibrium Xenon. Baron concentration is at 900 ppe.
- A : r eactor trip occurs. Assuming ALL rods trips what is the approxi-cate shutdown margin immediately after the trip?
a.' '9.04
'b. '8.04
- c. '7.0%
- d. '10.07 (1.0) 4 Q UES TION 1 11 (1.00)
.The reactor is at equilibrium Xenon. Baron .- 890 ppm, rods are in manual with Tave on program and the turbine is loaded to 475 MW.
Tur bi ne load suddenly reduces to 360 MWs but the steam dump falls to activate.- Assuming no protective action occurs, what will be the new steady state reactor power and Tayg?
- a. Reactor power 504, Tayg 587 F b.' Reactor power 50%, Tava 594 F
- c. Reactor power 38%, Tavo $87 F
- d. Reactor power 38%, Tavg 572 F
- o. Reactor. power 38%, Tavg 594 F (1.0)
. .-- . - . - - . . .- .. -. -- . . - . -. .x. ,_ -_-. . . -
la ERINCIELE3_DE_SUCLEAE_EDMER_ELAtil_QEERAIIQMa PAGE 6 IMER5001BA51C3a_dEAI_IRAMSEER_AHQ_ELUIQ_ELOW QUESTION- 1 12 (1 00)
Which Illustration below represents redistribution in the reactor ct 1005 power EOL7 ,, , , , __
ico%... -_
- a. g
- b. n' g .
2 Peak
\-
$ i E 50% -
~ Peak S0W-b b
,,,,,,__ *A,iAt --
- c. ioo % . ._ _ _ _ d.
ll 2
[
\ \
2
{ Peak I 50%- Peaks I 50%<
E ) 3 (l.0) o o hA xtAL " kAxlAL QUESTION 1 13 (1 00)
Which statement below correctly describes what happens to Beta-bar-ef f as the core ages?
- a. Its value decreases causing the reactor to respond faster to reactivity changes.
- b. Its value decreases causing the reactor to respond slower, to reactivity changes.
- c. Its value increases causing the reactor to respond slower to reactivity changes.
de 'Its value Increases causing the reactor to respond faster to reactivity changes. (1 0)
- La__ERINCIELE3_QE_BMCLEAR_EDMER_ELAHI_QP.EEAIIQ3a. PAGE 7 IMER50DIBA51Cla._ME AI_ IRAN S EER_ ABD_ELUIQ_ELQM QUESTION 1.14 (1.00)
Which of the following is the correct defini tion f or Doppler
'Tcoperature Coefficent?
c a. A change in fuel temperature due to a change in reactivity.
- b. A change in reactivi ty due to a change In the temperature of the -
fuele
- c. A change in reactivi ty due to a change In reactor power.
- d. A change in reactor power due to a change in fuel temperature. (1.0)
QUESTION 1.15 (1 00)
Hcw does differential boron worth vary as the core agest
- c. It initially increases then decreases.
b .1 It continually decreases.
- c. It continually increases.
- d. It initially decreases the increases. (1 0)
QUESTION 1.16 (1.00)
-10 With reactor power at 1*10 amps, a SUR of .7 DPM is established.
-6 Using 1*10 amps as the POAHs how long will it take the power to roach the PDAHT
- a. 2 8 minutes'
- b. 5.7 minutes
- c. 28.81 minutes
- d. 4.29 minutes (1 0) i l
1
- 1 u t81MCIELEl_DE_HUCLEAE_EDEER_ELAMI_QEERAIIDMa, PAGE 8 IMER 5DDIBA51Cla._UE AI_IBAM1EER _ AND_ELu1D_E L Q M QUESTION 1.17 (1.00) )
l A reactor is presently shutdown with a Keff of .90 and source ;
range counts at 10 cas. The oper ator inserts reactivity until l tho source range reads 65 cps. What is the new Keff?
- a. 0.95 .
l
- b. 0.98 )
- c. 0.90 !
- d. Insufficient data to calculate (1 0) l l
l QUESTION 1.18 (1.00)
As the core agess delta I at 100% equilbrium Xenons
- a. remains the same.
- b. becomes more negative due to more negative MTC. l
- c. :becomes less negative due to samarium.
- d. becomes less negative due to redistribution. (1.0) l 4
- La__ERIECIELE1_DE DUCLEAE_EDMER_EL&EI_QEERAIID$t PAGE 9 ItiERBDDIB&BICSa._tlEAI_IRAMSEER_AED ELu1R_ELQlf QUES TION 1.19 (2.00)
Mct ch each ter m in col umn A to its proper definition using the d3finition in column B.
COL UMN A COLUMN B .
- a. Heat flux 1. The process of heat transfer involving movement or circulation of fluid.
- b. DNBR
- 2. This takes pl ace when heat flux is high
- c. Nucle ate boi ling and steam is formed in a rel atively f lat and large. bubble.
de ' Film boiling
- 3. The ratio of the heat flux at a heated
- o. Convection surf ace th at will caus e DN8 to the actual o heat flux present.-
- f. Con du c t i on
. 4. The process of heat transfer invol ving mol ecu l ar vi bration.
- 5. This occurs when small steam bubbles are being f ormed on a heated surf ace and ar e released into the flow str eam.
- 6. BTU /hr per square f oot of heating surf ace.
CO.33 e ach ) (2.0)
QUESTION 1.20 (1.00)
The reactor is oper ati ng at 907 of rated thermal power with Tave on program and equi li brium Xenon. The secondary instruments indicate steam pressure at 805 ps ig. The computer shows f eedwater temperature at 440 F. Negate any energy added by th e RC P's and assume all l i quids at saturation. Calculate the mass flowrate in the secondary sy s tem .
6-
- a. 3.21*10 lbm/hr 6
- b. 10.96*10 lbm/hr-6
- c. 12 17*10 lbm/hr 6
- d. 3.57*10 lbm/hr (1.0) 4
- , , ,,,,,,,,----,,,,,,---a-
., - , , - - - - - - - , ,--w-.--n.,-%--~-yn-vew---,---~-.--,-,,,--m ---w-~--a ,e$<
% _ERIECIELEl DE_EMCLEAR EDidER_ELABI_QEEIL&ILQ1a. PAGE 10 LHEREQDIEA51 Cit MEAI_lRAHSEER_AHQ_ELu1Q ELat QUESTION 'l.21 (1.50)
AL centri fu gal char g i ng pump is running with its discharge va l we P Cr ti a l ly open. Indicate the direction (I NC RE A S E, D EC R E A S E s or REMAIN THE SAME) that each parameter will shif t if the discharge v alve is opened f ur the r.
- a. Discharge flow
- b. Pump discharge pressure
-c. Motor powe r
- d. Motor amps
- o. Available NPSH to pump C0.3 each) (1.5) 3UESTION 1.22 (1.00)
Steam exi ting th e HP tur bine is a t 785 psig, 90% qu ali t y. Steam entering the LP turbi ne is superhea ted to 100 F. W h at is the enthalpy change of the s team ?
- a. 154 BTU /ibe
- b. . 705 BTU /lba
- c. 85 BTU /lba
- d. 140 BTU /lbe (1. 0 )
i i
r -
.t- --- ,----n -,
- Za__ELAMI QSSISH_IMCLuDING_1AEEII ADQ.EBERGESCI_SISIEH1 PAGE 11 QUESTION 2.01. -(1.00)
The Reactor Vessel Head Vent System (RVHVS) di sch arge to the s
- a. Con t ai nm en t.
- b. RCDT. *
- c. W as te Ho ldup Ta nk.
- d. PRT. (1.0)
QUESTION 2.02 (1 00)
What installed plant equipment is used to detect leakage through the RVHVS isolation valves during normal plant operations?
- a. -Containment pressure / humidity detectors.
- b. Containment sump level monitors.
- c. Acoustic Leak Monitor System (ALMS).
- d. Radiation Monitor System. (1.0)
QUESTION 2.03 (1.00)
Th3 purpose of the Reactor Coolant Pump (RCP) Seal standpipe is to provide as
- a. means of collecting #3 seal leakoff.
- b. lubricating water supply for #3 seal.
- c. collection point for the #2 seal leskoff.
- d. pressure head for #3 seal. (1.0) 1
Za__ELAHI_DE11EH_lHCLu2INE_1AEEII AHQ_EBEREEMCI_1XIIEBS PAGE 12
-QUESTION 2.04 (1.00)
ROcctor Coolant Pump #3 is started during shutdown after a seal
-rCPlacement. After operating for approximately 20 minutes the following is observed.
- 1. #1 seal delta-P >4008
- 2. Standpipe low ' level
- 3. #1 seal leakoff has increased. '
ASSUME:
- 1. Plant pressure is at 400#
- 2. Seal injection at 6 ops.
Which of the f.ollowing is a probable cause for the abnormal Indications?-
- a. 83 seal failure.
- b. VCT pressure is low.
- c. #1 seal bypass is open.
- d. RCDT pressure has increased. (1 0)
QUESTION 2.05 (1.00)
'The purpose of the interlock that prevents the letdown isolation valves from opening or shutting unless all three ori fice isolation val ves are shut is to prevents
- a. exceeding design flow rates of the demineralizers.
- b. excessive heatup rates across the regen. heat exchanger.
- c. excessive pressure on the shell of the regen. heat exchanger.
'd . unnecessary lifting of relief valves downstream of orifices. (1.0)
la__ELAMI_REllSH_1HCLualHE_1&EEII_AND_E5EEEH CI_SISIERS PAGE 13 QUESTION 2.06 (1.00)
Which of the following is correct concerning the operation of the
- Co cen tr i f uga l char ging pump during a Blackout? ASSUME that tho "C" pump is in standby at the time of the Blackout.
- a. "C" Pump .will load onto the selected emergency bus af ter the ,
sequencer-has timed out. .
- b. "C" pump wil l be lockedout to prevent overloading the diesel gen er ator. i
- c. "C" pump will start on the Blackout if the "A" or "B" pump fails to start.
- d. - "C" pump must be manually started i f either " A" or "8" pump i s out of service for maintenance. (1 0) l
. QUESTION 2.07 (1.00)
The pressurizer surge line connects to loop ______ and the spray lines connect to loops ____ and ____ a
- a. C, A&B.
- b. B, A&B.
- c. Cr'OLC.
de As AGC. (1.0)
QUESTION 2.08 (1.00)
What are TWO reasons f or maintaining a minimum spray bypass flow-
.to the pressurizer? Choose only one of the following.
.a. - 1. Prevent excessive cooling to the surge line.
- 2. reduce the delta pressure across the spray valves.
- b. 1. reduce thermal shock to the spray nozzle.
- 2. ensure that the backup heaters cycle on.
- c. 1. prevent excessive cooling to spray line.
. 2. equalize boron in pressurizer with the RCS.
'd. 1. minimize stress to the surge line thermal sleeve.
- 2. remove gases from the RCS. (1 0)
L m- -,-r,<,mm. - _ , - - , _ . - . . . . , , _ . _ _ . , , _ . _ , , . . . - , , . . _ _ _ , . - , ---,,_,-.m,_-_-,,.._...,__-_._.._.,,,--,-_-.,
"Es- ELAHI-QE31EM INCLUQ1HS 1&EEII AHQ EREEEENC1_313IE53 PAGE 14 l
QUESTION 2.09- (2.50)
Mttch the component of Column A to the automatic action in column B ,
which occurs during a Safety Injection. ASSUME all system lineups l aro normal. i COLUMN A COLUMN 8
- s. -Cooling water supply to reactor 1. Started.
bldg. cool er s.
- 2. Tripp ed and started
- b. RPI coolin g unit cooling wa te r in hi gh speed.
supply.
3.i S hi f ted to Service
- o. . S ervi ce water tr a vel ing screens. Water.
de Running Service Water Pump. 4. Isolated.
_ c. Ser vi ce Wa ter Booste r Pump.- 5. Runs continuously.
6.. Tripped.
[05 each) (2.5)
Q UES TION 2 10 (1.00)
The Sump-to-RHR suction valves automatic swi tchover during SI can ,
bo defeated by the operator perf orming which action belowf
- a. Reset SI signal e- ee-tre! 5 ::.- d e -
- b. Roset reactor tri p breakers.
- c. Reset the RWST 2/4 logic I vl L o-L o b i st abl e.
od . Ros et the master SI signal er ee-tre! 5ee'd 4: __ (1.0)
QUES T ION ~ 2.11 (1 00)
Tho Condensate Storage Tank minimum water volume of 172,k00 gallons is sufficient to maintain the plant in hot standby for how many hourst
- a. 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />
- b. 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br />
- c. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> de 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (1.0)
i
. a 1
'EnELMLDE11EtLitiCLUD15E_S&EEII MLil1EMEliCL111IE51 PAGE 15 4
QUESTION 2.12 (1.00)
What prowl des t he means of shutting the flow control valves f or tho emergency f eedwater pumps should the normal air su pply f a ll? .
l
- a. Ni tro gen pr essure backup. j
- b. Air accu mu la tor s. .
i
- c. E lectr ic motor ope rator.
-d. Hydraulic oil supply. (1 0 l
l Q UES TION 2 13 (1 00)
Which action below i s t aken to pr otec t the Emergenc y Feedwater pumps from a runout condition?
- a. Flow orifices in the discharge line of each pump.
- b. Trip of motor breaker and/or turbine throttle valve.
- c. Automatic closure of the flow control valves. l
- d. Recirculation of the excess flow to CST. (1.0) i QUESTION 2.14 (1.00)
When ALL the Emergency Feedwater pumps are used for normal plant startups what percentageoof full load can be maintained?
\
- a. 1-2 l
- b. 3-4
- c. 5-6
- d. 7-6 (1 0)
ta__ELAMI_QESISH_IMCLuDIBE_1&EEIL&50_E5EREERCI 111IEH1 PAGE 16
-QUESTION 2 15 (1.50)
Which 3 o f the f ollowing d ie se t engin e/ gen er ator s hutdowns ar e en abl ed dur i ng . an emer ge ncy start of the diesel?
C.' High lube oli temperature (175 F).
- b. Hi gh Jacke t cool an t temp e r atu r e (195 F ).. .
- c. D iese l .ove r soned.
- d. Generator diff erenti al current.
- o. Genera tor overcurrent.
- f. Negative phase sequence.
- g. D ie se l low lube oil *
- h. Reverse power.
I. Generator ground. [ 0.5 each] ( 1. 5 )
QUESTIDN 2 16 (1.00)
Durin g the r ecov ery phase of a pl ant Bl ackout t he norma l of f-si te powor- suppli es are placed back in service. What action must the eparator take to insert speed droop control in the diesel governor system before paralleling the power sources?
- a. --Depress Emergency Start Reset and s elect parallel operation on the Vol tage Regul ator switch. ;
b.' Selec t parallel operation on the Voltage Regulator switch and turn on the synchroni zin g s cope.
- c. Depre ss Emergency St art R oset and Test Start pushbfttons.
- d. Roset Blackout sequencer and depress the Exciter Reset switch. (1.0) l
r
- Za__ELAMI_QESIGH_IRCLUDING S&EEII_AHQ_EREEEENC1 1111151 PAGE 17 QUESTION 2.17 (1 00)
Tho EDG can oper ate loaded at 5100 KW and rated powerfactor fort
- a. 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />.
- b. 30 minutes.
- o. 7 d ay s .
- d. Continuously. (1.0)
QUESTION 2 18 (1.00)
Tno "Emer gency Start Override" pushbutton on the Emergency Diesel ctntrol panel allows the operator tot
- a. tske control of D/G and shutdown D/G af ter an emergency start.
be reinstate D/G trips lockedout by an automatic or manual start.
- c. Ini tiate a D/G start when support systems are not functioning.
de reset the D/G emergency start circuits. (1.0)
~ QUESTION 2.19 (1.00)
The DOMINANT power supply path for the 120 VAC vital busses is front
- a. 480 VAC vital busses thru inverter.
- b. 480 VAC vital busses thru regulated stepdown transformers.
- c. 125 VOC power from panels DPN 1HA and 1HB.
2 1
- d. 7.2 KVAC v i tal busses thru regulated stepdown transformers to )
120 VAC. (1 0) l l
't..
2n__EL&BI_DE11SH_1HCLUDING_S&EEII_&MA_E51RSENGI_11SIE51 PAGE 18 QUES T ION 2.20 (1.00)
HWhi ch tr i p of the 7.2 KV safeguards busses IDA and IDS will cause all l o a ds to be dropped and then seqenced back onto the bust as Overcurrent
.b. Under yof tage .
- c. Phase Olfferential
- d. Ground fault (1 0)
QUESTION 2.21 (1.00)
?
- Which TWO conditions will cause the automatic shutting of the feed Isolation valves in or der to prevent water hausert a .~ Low steam pressur e and. low S/G le ve l .
- b. Low f eed t emper atur e and low S/G level .
cf Low feed flow and low feed te mper a t ure .
'd. Low steam pressure and low feed flow. (1.0)
QUESTION 2.22 (1.00)
During the star tup of the secondary system the f eedwater piping must bo warmed up to a point Just upstream of the f eed Isolation valves.
This warm up flow must be maintained at 250 ope f o r a t l e a s t _____
oinutes. Then a reverse flush is done at 70 gpm and 400 F for
_____ m in ute s.
- a. 40s 30
- b. 30s 20
- c. 20, 40
- d. 15s 40 (1 0)
r
'Ia _EL&Hl_DESISB.DCLUQ D S_S&EEI1_At1D_EEEEEENCI 11SIE53 PAGE 19 QUESTION 2.23 (1.00)
A MULTIPLE LINE-OF-DEFENSE concep t is us ed i n the design of a nuclear power plant in order to minimize risk to the public.
- Match the DEFENSE in column B to the LEVEL OF DEFENSE in column A.
C OL UMN A COLUMN 8 .
- a. ' FIRST 1. Reactor Protection and Logic System.
- b. SECON D
- 2. Eng ineering S af egu ar ds
- c. THIRD Systems (ES F).
- 3. Proper desi gn, construction techn i qu e, an d m at er i a l .
[0.33 e ach3 (1.0)
r-I g ,,1ggIgggggJs.ggg.cggIggLg PAGE 20 l l
l l
i QUESTION 3.01 (1.00)
Bscause the Steam Generator PORV's are part of the Steam Dump System they must shi f t between control systems when needed.
H:w is this accomplished? j
- o. Two additional solenlods in the control air path.
- b. Two relays transfer the input signals to the I/P converters.
- c. One solenlod and one relay tr ansf er the air and control signal.
- d. Two independent controllers supply parallel signals with the auctioneered high signal controlling. (1 0)
QUESTION 3.02 (2 00)
Match the statement in Column A to the proper Steam Dump mode in Column B.
COLUMN A COLUMN 8
==
- a. Operates Steam Dump with a 1. Turbine trip 5 F deviation (Tavg-Tref).
- 2. Steam pressure be Has two independent oper ating sections. 3. Load rejection
- c. Controls only the condenser steam 4. Tavg dump bank of valves.
- d. Operates steam dump with no dead band (Tavg-Tref). [0.5 each) (2.0)
QUESTION 3.03 (1 00)
Th3 THREE input signals to the Steam Generator Water Level Control aros 1
- a. T av g, compensated feed flow, uncompensated steam flow.
- b. Feed flow, compensated steam flow, water level error.
- c. Compensated feed flow, water level, compensated steam flow.
de Uncompensated feed flow, compensated steam flow, water level. (1 0) ,
l i
l
- .~ - . - - . _-
l l
I
'3a _1M3IRU5EMI1_AND_GDHIRDL3 PAGE 21 1
QUESTION 3.04 (1.00)
Which of the following is used to create the level programming signal fer the S team Generator Water Level Control System?
l C. Average NI power.
- b. Auctioneered low NI power. .
l
- o. Power range NI 44.
- d. Auctioneered hl NI power. (1 0)
QUESTION 3.05 (1 00)
Csnsidering only the Steam Generator Level Control System, what wguld be the response of the INITIAL feedwater flow, if the 1
ccntrolling S/G pressure transmitter failed LOW during 50% power
-cperations?
- s. The flow would decrease due to the loss of the steam pressure input to the steam flow signal.
- b. The flow would remain the same due to the steam pressure not ,
affecting the steam flow. I
- c. The flow would increase due to the steam pr essure input to the feed control valve position.
I
- d. The flow would increase due to the loss of steam pressure -1 input to the steam flow signal. (1.0) i I
l l
QUESTION .3.06 (1.00)
Prossurizer level program is based on: 1 l
- a. Power Range NI 44.
- b. Tavg.
- c. Tref.
- d. Auctioneered High Power Range. (1 0)
la._1811RMBENIS_AND COMIRDL3 PAGE 22 QUESTION 3.07 (1.00)
To- prevent the possibilty of excessive and unnecessary depressur-lactions, the CO PS s e tp o i n ts ar e _______________ t o avo i d o pe n i n g o sccond PORV during less severe tr an s i e nt s.
- a. alternated )q be blocked et c
O. staggered i d. disabled (1.0)
QUESTION 3.08 (1.00) .
Which of the following is TRUE concerning the operation of the COPST
- c. The COPS use auctioneered low Tcold for temperature input to are them and auctioneered low Thot to disable thier operation.
- b. The COPS will cause an alarm if either PORV block valve is st,ut below 300 F. j
- c. The COPS uses a olstable actuated temperatu i n te r l ock from the opposite channel to prevent inadver ent actuation.
I
- d. The operation of either COPS valve will protect the plant from a pressure excursion due to both a heat input transient and a :
mass input transient. (1.0) l 1
-QUESTION 3.09 (1.00)
Th3 controlling pressurizer level channel (459) falls high during 1004 power operation. Assuming NO operator action is taken.
Which of the following best describes the response of the plantt
- a. Charging flow goes to minimum, pzr level de cr eas es, letdown
' Isolates and the plant continues to operate at the same power.
- b. Charging flow goes to minimums pzr level de cr e ases, l e tdo wn
- Isolates and the plant trips on high par level.
c.- Charging flow goes the maximums pzr level increases and the plant trips on high Pzr level.
de Charging flow remains the sames pzr level increases due to letdown isolating and the plant trips on high pzr level. (1 0)
_._....._._..._-_.______.-___m.____._... _-
l
'as__1811&WHEMI1_ANE COMIRQL1 PAGE 23 l
l QUESTION 3 10 (1.00)
Which a f the following core parameters does the OT delta T protec tive circui t pr event exceed ing?
a.. Pow er dens i ty
- b. Depar ture f r om Nucleate Bolli ng . l
- e. Total core power
- d. R edis tr i bu t i on (1.0)
Q UES TION . 3.11 (1.00)
Tho OP del ta T setpoint is a calculated value determined by whi ch of the following methods?
~
- a. 109% full power delta T minus correction proportional to rate of inc re ase o f Tavge minus correc tion proportion al to v ari ation of del ta T abov e fu ll load value.
- b. 109% full power delta T minus. correction proportional to rate of incre as e o f Thot, minus correction proportional to varation of Thot above fu ll l oad Thot
- c. 109% fu ll power delta T minus correction pr oportional to rate of increase of Tavg, minus correction proporti onal to varation of Tav g above f ull load Tavg.
de 109% full power del ta T, minus correction propor tional to rate of increase of delta Ts minus correction proportional to verlation of delta T above full load. (1 0)
'3a._L53ILDEEEL1_AND CDMI80Ls 'PAGE 24 1
i QUESTION 3. 12 (2.00)
Indicate whether the f ollowing statements concerning the Nuclear Instrum ent System ar e TRUE or FALSE.
- a. -The source range instrument uses a fission chamber f or det ecti ng neutr ons. - (0.5) ,
- b. When an intermediate range detector is over' c ompensa ted the meter r eadi ng will be l o we r th an th e actu a l neutr on f l ux le vel . (0.5)
- c. The power range CHANNEL CURRENT COMPARATOR ( l oca te d in the comparator and rate drawer) outputs an alarm i f an y ch ann el devis tes by more th an 2% fu ll power f rom any other channel. (0.5)
- d. Power range channel N44 i s us ed by the reac tor protection and logic system for input to the steam generator low level !
and low-low l ev el setpoints. (0 5) !
QUESTION 3 13 (1.00) l The f unction o f the Reactor Protec tion Systen i s to pr event damage to the re actor core and subsequen t rel ease o f materi al to the gcncr al pu blic. What are the FIRST TWO barri ers to the release fission products?
- s. Pr i mar y coolant piping and steam generator tu be s .
- b. Fue l c ladd ing and primary coo l ant pi pi ng.
- c. Pr imary coolant piping and reactor containment.
- d. - Fuel cladding and reactor containment. (1 0)
'3a__LN3IRygEgli_ANQ_CDHIRQL1 PAGE 25
' Q UE STION 3.14 (3.00)
-Match the conditions of column A to the permissive relay or control rol ay -o f co lumn 8.
COLUMN A COLUMN 8 P-10
- a. Blocks rod motion if 1/4 channels 1. .
of power range reaches 103%
- b. Automaticly blocks k# reactor trips when r eactor power and turbine power are <104. 3. C-2
- c. Block rod wi thd raw al i f intermediate 4. P-7 range powe r is >204 equi v alent and itself is not blocked. 5. C-1
- d. 3/4 powe r ranges below 38 % power. 6. P-12
- e. Trips all feed pumpss Is ol a te s 7. P-11 feedwater and trips turbine if 2/3 Hi-Hl.S/G lev els >82 44. 8. P-8 QUESTION 3.15 (1 00)
Tho pr imary -pur pose of the time delay i n t ri p p i ng t he mat iunrt /4420 aftcr tripping the reactor is tot 35*
as Reduce arcing of the generato r output br ea ker s.
- b. : Prevent over speeding of the m ain tur bi ne. l C. Minimize the conse quences o f a reactor tr ip.
de Remove excess steam from the moisture /seperator reheaters. (1.0) l l
Q UES TION 3.16 (1.00)
On an automatic reactor trips t he tr i p b r e ake r s i n te rr up t __________
-pecor to the rod control power cabinets by de-energizing the
___ __________ c o l i o f t h e tr i p br e a ke r s .
ei a. 260 VDC, trip
- b. 120 VDCs s hu nt (AV ($ l
- c. 260 V ACs s hu nt, tv cdl s~ d . 120 VACs trip (1.0)
3.__INSIRUDEBJ3_A50 CQMIRQL1 PAGE 26 ,
QUES TION 3.17 ( 1. 00)
Which of the following s t at em en ts is TRUE conce rning the core ccoling moni tor?
- a. Shows how close the reactor coolant is to bolling.
- b. Yellow status li ght Indicates tha t the cool ant i s at s aturation. '
- c. All lights will f l as h i f the ' ALARM ACKNOWLEDGE' pushbutton .
has been pushed.
- d. Uses the lowest temperature and the lowest pressure to determine the displayed numerical v al ue o f su b-coo li ng. (1.0) l QUESTION 3.18 (1 00)
The Reactor Vessel Level Indication System uses what type of instrument to determine the water Inventory of the vesself
- a. RTD's i
- b. Thermocouples
! c. Fission chambers de De l ta P cells (1.0) l 4 3UESTION 3 19 (1 00)
Tho resultant Indication of the Reactor Vessel Level In di c at i on
~ Sy s te m is corrected for adverse c onta inment condi tions by
- a. Sensing line temper aturess wide range Thot, and RCS pressures.
- b. Containment air te mperature ss containment pressure and
. tr ansmi tter l oc a t i on.
- c. Reactor vessel wall tempe ra tu res RCS Taves and loop pressure.
- d. Containment temperatures containment pressure and operator use of a graph. (1.0)
r '. l
- 1r.__IH11RUBENIS_AHQ.CONIRDL1 PAGE 27 i
-QUESTION 3.20 (1.00)
Dur ing power operations the Steam Generator water level as indicated is en a programmed r amp over part of the power range. How does this affcet the total mass inventory in the Steam Generators 7
- a. Mass of water decreases and steam mass increases.
- b. . Mass of water decreases and steam mass remains the same.
- c. Mass of water increases and steam mass decreases.
- d. Mass of water increases and steam increases. (1.0)
QUESTION 3 21 (1 00)
Listed below are all of the condi tions which will start the Ecorgency Feedwater Pumps.
- 1. -2/3 low-low S/G level f rom any S/G
- 2. 2/3 low-low S/G level from 2/3 S/G
- 3. All main feed pumps trip
- 4. Blackout
- 5. . Safety Injection Which conditions below will cause only the TURBINE-DRIVEN pump to start?
- a. I and 5
- b. 3 and 4
- c. 2 and 4
- d. 2 and 5 (1.0) s
- . . _ . _ . - , . - _ ~ . _ . - _ . - . - - - - _ . _ , _ , _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ . . _ . _ _ _ _ . _ . _ _ _ . . . _ - ,
'ia__REQCEQQRE1_:_8085&La_A18085Aka_EBEREENGl_AHQ PAGE 28 RADIDLDSICAL_GQ8IRQL l
QUESTION 4.01 (1.00)
Pricr to operating Reactor Coolant Pumps the minimum seal flow should be _____ sps and VCT pressure should be a minimum of
_____ Psig.
- a. 1.Os 10 *
. j
- b. 0.2s 15
- c. 0.5s 30
- d. 0.75, 20 (1.0)
QUES TION 4.02 (1 00)
During normal power operations, CVCS letdown flow is lost. What itacdiste action must be taken in regard to charging flow and why must this action be taken?
- a. FCV 122 must be closed to prevent thermal shock to the reactor coolant system. *l
- b. FCV 122 must be opened to prevent overpressurization of the regenerati ve heat exchanger. ,
l c. FCV 122 must be closed to direct all water to the reactor b coolant pump seals.
- d. FCV 122 must be opened to maintain proper charging flow to the reactor coolant system. (1.0)
QUESTION 4.03 (1.00)
What is the maximum percent by volume oxygen concentration in the -
CVCS VCT during normal power oper ation and why is this limit imposedt a.- 3%s to avoid flammable alxtures in the gas space.
~
- b. 74, to allow controlled recombination of the gas mixture.
- c. 5%s to avoid explosive mixtures in the gas space.
- d. 44s to ensure non-explosive mixtures in the gas space. (1.0) g .
'iwERGCEDURE1 - 5015&La_AR50E5ALa_15EEEEBCI_AMD PAGE 29 RADIDLDGICAL_COMIRDL QUESTION 4.04 (1.00)
It is des ired to perform a RCS dilution to a cr itical boron c:ncentration with the pl ant in hot-standby and Xenon f ree etnditions with all rods inserted. Prior to the RCS dilution, th3 operator must haves
- a. 4 Shu tdown bank s wi thdrawn.
- b. 3 Shutdown b anks wi thdr awn.
- c. 2 Shutdown banks wi thdrawn. l
- d. 1 Shutdown b ank w i t hdr awn. (1.0)
Q UEST ION 4 05 (1 00)
What ar e FOUR Indications used to monitor pl ant cooldown during Natur al Ci rcu lat ion according to E0P-1.3 7
According to E0P-1.3 " Natural Circulation" procedures what is tho operator's first priority?
- a. Start RCP A or C.
- b. Borate to cold shutdown conditions.
O Establich controlled cooldowh rate.
d6' Maintain subcooling margin > 30 F. (1.0) 1
~;,
s
-. . _ _ . _ n
%__ RS QCED UREl_:_BDA B A L a_ AABQ R B A La _ERE RG ENCL.4M Q PAGE 30 RAQ1DLDEICAL_CQHIRQL QUESTION 4.07 (1.00)
Af tcr initiating the " Loss o f Cool ant" procedur e E0P-2.0s at I what decreasing pressure must the Reactor Coolant Pumps be steppcdf I
- a. 1550 psig *
- b. 1380 psig
- c. 1400 psig
- d. 1650 psig (1.0)
QUESTION 4.08 (1.50)
Which of the following conditions would require re-Initiation of safety injection according to E0P-1.2 "Saf ety Injection Termination"?
RCS PRESSURE SUSC00 LING PZR LEVEL %
- a. stable 25 15
- b. Increasing 40 3 ce decreasing 30 10 de Stable 35 5 (1 5)
QUESTION 4.09 (1.00)
During boron concentration changes to the RCS at power the operator cust observe one of two effects caused by the boron change. What aro the TWO effects?
- a. Control bank motion and changes in RCS Tavg.
- b. Spray valve actuation and NI power change.
- c. Control bank motion and NI power changes.
de Changes in RCS Tave and Spray valve actuation. (1 0)
- ka_.tRDCEQll1EL HQ15&La.AAMQERALa_EBEREEHCL&3Q PAGE 31 RARIDLDGICAL_COMIRQL QUESTION 4 10 (1.00)
Pricr to increasing Tayg from mode 5s your heatup procedure (GDP-1) gives the option NOT to withdraw shutdown banks. Wh at condition cust exist prior to taking this option?
- a. Boron samples taken and all trip breakers open. *
I
- c. The core is not Xenon free.
de RCS borated to cold shutdown conditions. (1 0)
QUESTION 4 11 (1.00)
Which ONE of the following would be the BEST Indication of a rcactor tript
- a. Reactor Trip annunclator. l
- b. Rod bottom lites out.
- c. Auxillary feed pumps on.
- d. Power level decr easi ng r apidly. (1.0) l QUESTION 4.12 (1.00) l l R30alcul ation of the Estimated Critical Condition must be done if L tho startup has not occurred wi th in _____ hours of des i gna ted time.
- a. Two i
- b. One
- 0. F ou r de Eight (1.0) l l
r 1 l
'As..ELEC1QUEE1_=_EQR5Akt AASDRBALa_EBEREENCI.&5Q PAGE 32 LAaloLOGICAL_CDMIRDL l l
3UESTION 4 13 (1.00)
If cctual critical rod position is below the low-low rod Insertion iloits GOP-3 states that you shall t
- a. Shutdown the reactors recalcul ate ECC. ,
- b. Continue with startups recalcul at e ECC.
- c. Emer gency Bora te, shutdown the reactors recalculate ECC.
- d. Tr ip the reactor, Emergency Bor at e, recalculate EC C. (1.0)
Q UES TION 4.14 (2.50)
Match the terms in column A to the values in column 8 for the rcdiation _ exposure guidelines. Assume whole body dose unl ess otherwise stated. CAUTION: Some answers could be used more than cnco.
COLUMN A COL'JMN B
- a. NRC limits /qtr 1. 05 REM
- b. V. C. Summer limits /qtr 2. 1.25 REM ,
- c. NRC pregant women ilmit/ gestation 3. 1.0 REM
- d. NRC general public limit / year 4. 0.75 REM
r a
h -_ERQCEDUEE1 =_HQAHALa._AAMDERALa._EBEREENCX_ANQ PAGE 33 1ADIDLDGICAL CQHIRDL QUESTION 4.15 (1 00)
Hic is an occurrance of Pressurizer temperature to spray temperature oncoeding the 320 F limit documentedt
- a. Licence Event Report filed.
- b. Reactor Engineer notifled immediately.
- c. Logged in operator's log.
- d. No action required. (1.0)
QUESTION 4.16 (1.00)
Which statement is TRUE concerning the term Maximum Permissable Concentration?
- a. The airborne limits for noble gases are based on internal doses.
- b. MPC-hour exposure will be calculated for workers exposed to airborne radioactivity including noble gases.
- c. Radiation Warning signs will be posted for airborne radioactivity in areas where the weighted MPC value is > 0.25.
- d. Personnel shall not knowingly be exposed to any airborne radioactivity with weighted MPC value greater than 15 without respiratory protection. (1 0)
b, 1.__280CEDURE1_=_HORMALa AAHOREALa_EBEREEHC1_AMQ PAGE 34 l RAQ10LQs1EAL_CDHIRQL Q UES TION 4 17 (3.00)
Match the condition in column A to the action required in column 8.
COLUMN A i
- s. Shutdown Margin in Question. 1. Reduce plant load. '
- b. 2/3 lo Steamline press. < 6758 2. Roset mech. over-speed.
c.- Emergency Desiel trip during SI
- 3. Emer gency bor ate.
d.' Feed' booster pump trip
- 4. Place Rods in O. PZR pressure control channel f alls manJal.
High.
- 5. Safety inject. ,
- f. First stage turbine pressure f ails High. 6. Close PCV-444.
C0.5 each) (3.0)
QUESTION 4.18 (2.00)
TRUE or F ALSE concer ni ng a dr opped rod. ,
, s. The rod control s tar tup re se t swi tch i s depr ess ed pr i or to the r ecovery of a dr opped rod. (0 5)
.. b . Tur bi ne load is decreased to 80% power and then Tavg-Tref are equalized to +/- 3 to 5 F. (0.5)
'c. Af ter recovery the rod which was dropped is exercised to v er i f y oper ab il i ty. (0. 5 ) .
- d. .The step counter is manually reset to zero f or the recovery of the dropped rod. (0.5)
(-
's a__ ERQC ED U RE1_=_d QAd els_ AB BOR BA L s_EBE E E ENC 1_&1R PAGE 35 R&RIOLDEICAL_CONIROL QUESTION 4.19 (1.00)
In addi ti on to the FOUR cond i tion s fo r termi n ating a Reactor Hoad vent belows what is the FIFTH criteri a?
- 1. S u b coo l i ng < 30 F
- 2. 'PZR level <171 '
- 3. RCS pressure decreased by 200 psi
- 4. Ves s e l level stable and normal
- s. Conta inment pressure increases by 0 5 psi.
- b. Prede te rmined venting per iod is met.
- c. Vent line fi sw suddenly increases.
- d. PRT pressure increases to 75 psig. (1.0)
- Q UESTION 4.20 (1.00)
The Cri tical S af ety Functions have an order of i m po r t an ce asscci ated wit h them. Order the Safety Functions below in d0cending importance.
t
- 1. Containment
- 2. Heatsink
- 3. Suberiticality
- 4. Inventory
- 5. Core cooling
- 6. Integrity- (1 0) i 1
~ ,m. ,--v_.. ..m_, , , ____. ,__ ,_ , _ , , _ , , ,
P l
i
- Is__ERIECIELE1_DE_SWCLEAE_EDEER_ELadI_DEERA110$a PAGE 36 IMEE50DINA51CSA_HEAI_ISAH1EER ASQ_ELulD_ELQW ANSWERS -- SUMMER -85/02/25-PICKER, B.
, aA ANSWER 1.01 (1.00) ,p) answert b. (1.3)
REFERENCE V. C. Summer Curve Books Fig. II-1 ANSWER 1.02 (1.00) answert b. (1 0)
, REFERENCE V. C. Summer Curve Book, Figures II-2as II-2c, II-1 ANSWER 1 03 (1.00) answert d. (0.5)
REF ERENCE VCS Curve Book, Fig. II-10 ANSWER 1.04 (1 00) answert a. (1.0)
REFERENCE V. C. Summer Reactor Theory, I-5.80 ANSWER 1.05 (1.00) answers c. (1.0) l REFERENCE V. C. Summer Heat Tr ansf ers Thermodynamics and Fluid Flows pp 226, 243 1 I
1.__E81HCIELE1_QE_HuCLEAE_EDEER_ELABI_DEERAIIDHg PAGE 37 IMEE5001HA51CSA_MEAI_IEANSEE8_AHQ_ELUIQ_ELQM
. ANS WE RS -- SUMPER -85/ 02/ 2 5- PIC KE R, B.
ANSWER 1 06 (1.00) ans wer s b. (1.0)
REFERENCE V. C. Summer HTFF pp 94- 96 A NS WER 1.07 (2.00)
- a. F al se .
- b. False.
- c. True.
- d. F al se . [ 0. 5 e a . ] (2.0)
R EF ER EN CE VCS Heat Trans fer an d F lui d F l ow, pp. 322-334.
ANSWER 1.08 (1.00) answers d. (1.0)
REFERENCE Wostinghouse Thermo-nydr aulic Principles and Applici ation to PWR, II, ch. 12 pp 21-26 ANSWER 1 09 (1 50)
- a. Reactor A (0 5) b.. Reactor B (0.5) l
- c. They will have the same rod height at 10 -8 amps. (0.5)
REFERENCE VCS Reactor Theory, pp 4.7-4.30, 5.36-5.50
- la__ERIMCIELE1_QE_HQCLEAE_EQMER_ELABI_ DEER &IlQHg PAGE 38 IMEE5001BA51 Cit _MEAI_1EAd1EER_ABD_ELu1Q_ELQM ANSWERS -- SUMMER -85/02/25-PICKERS B.
ANSWER 1.10 (1.00) answers b. (1 0)
REFERENCE VCS Curve Books II-9s II-12 ANSWER 1.11 (1.00)
O answers c. (1.#1 REFERENCE VCS Reactor Theory, p. I-5.26 ANSWER 1 12 (1.00) answers c. (1 0)
REFERENCE .
.VCS Reactor Theorys I-5.54 ANSWER 1.13 (1 00) answers a. (1.0)
REFERENCE VCS Reactor Theorys p. I-3.9 ANSWER 1 14 (1.00) answers b. (1 0)
REFERENCE VCS Reactor Theorys pp I-5.16 and I-5.26
--.__-_ _ _.~. _ _ . _ - - . - _ - _ . - - - - - _ . _ . _ _
. l
' l i__ ERINCI EL ES_ DE_SU C LE AE_EQW ER_EL A MI_ QEEE AIIGSt PAGE 39 IHERdQD1HABICla_dEAI_IRANSEER_AND_ELula_ELQM ANSWERS -- SUMMER -85/02/25-PICKER, B.
ANSWER 1.15 (1.00) anscers d. (1.0)
REFERENCE VCS Reactor Theory, p. I-5.35, Fig. 1-5.27 ANSWER 1.16 (1 00) answers b. (1 0)
REFERENCE VCS Reactor Theory, p. I-3.15 ANSWER 1.17 (1.00) answers b. - (1 01 REFERENCE VCS Reactor Theory, p. I-4.28 A NSWER 1.18 (1 00) answert d.. (1 0)
REFERENCE
-VCS Reactor Theory, Section 5, pp 52-54 ANSWER 1 19 (2.00)
- a. 6
- b. 3
- c. 5
- d. 2
- 0. 1
- f. 4 [0.33 each) (2.0)
~16._EElBCIELE1_QE_MuCLEAE_EQWEE_ELABl_QEER&IIQHa PAGE 40
-IDEE 50DXB&BIC3a_dEAI_IRAM1EER_ABD.ELMIQ_ELQM ANSWERS -- SUMMER -85/02/25-PICKER, B.
REF ERENCE VCS HTFF pp 99s 105, 123, 124, 228.
. ANSWER 1.20 (1 00) answers b. (1 0)
REFERENCE VCS HTFFs p. 179 ANSWER 1.21 (1 50)
- a. Increases
- b. decreases
- c. Increases
- d. increases 'O'3 E**b 0060
- e. decreases 0.00 ::05? (1 5)
REFERENCE VCS HTFF, S ect i o n 3-p a r t b-c h ap t e r is pp 324-329 ANSWER 1 22 (1.00) answert a. (1.0)
REFERENCE
-VCS HTFF pp 82-96) pp 143-144 l
i
- E s__ EL &dI_QE11GB_IUCL ualBG_SAEEL1_ AND_EEEEEENC1_ SIIIEMS PAGE 41 ANSWERS -- SUMMER -8 5 /02/25-PICK ER, B.
ANSWER 2.01 (1.00) a ns we r s d. (1 0)
R EF ER ENCE .
VCS Plant System Descriptions, AB-2s p. 26 A N SW ER 2 02 (1 00) answers c. (1.0)
REFERENCE VCS Pl ant Syst em Descr iptionss AB-2s p. 27 ANSWER 2.03 (1 00) answer s d. (1.0)
REFERENCE VCS Plant Sy s tem De sc r ip tion ss AB-4, p. 27 ANSWER 2 04 (1.00) answers b. (1.0)
R EF ER EN CE VCS Pl ant S ystem Des cr i ptions, AB-4s pp 29-31 ANSWER -2 05 (1.00) answers c. (1 0) 4 REFERENCE VCS Pl ant Syst em Descr iptions, AB-3s p. 10 and Revi ew Questions
"E6__EL&dI_QESIGH_lHCLUQ1HG 1AEEL1_AND_EdEEEENCI.SISIf 51 PAGE 42 A NS WE R S -- S UMM ER -85/02/25-PICKER, B.
A N S WER 2.06 (1.00) ans we r t b. (1.0)
R EF EREN CE VCS Plant System Des cr iptions, AB-3s pp 24s 25 ANSWER 2.07 (1.00) answer d. (1.0)
REFERENCE VCS Pl ant Syst em Descr iptions, AB-2, fi gur e AB-2. 3 ANSWER 2 08 (1 00) answer 8 c. (1.0)
REFERENCE VCS Plant System Descriptionss AB-2, pp 33-35 and Review Question #13.
ANSWER 2 09 (2.50)
- a. 3
- b. 4
- c. 5
- d. 2
- e. 1 Co.5 each) (2.5)
REFERENCE VCS Pl ant System Des cr i ptions, IB-5s pp 4s 6s 17 ANSWER 2 10 (1.00) answers a. (1 0)
REFERENCE VCS Plant System Descriptionss AB-10, pp 15, 16
e~
E a__ EL AHI_QESIGH_IUCLuQING_1&EEII_ AHQ_EMEREESC1_11SIEB1 PAGE 43 i
ANSWERS -- SUMMER -85/02/25-PICKER, B.
i ANSWER 2.11 (1.00) ansters a. (1.0)
REFERENCE -
VCS Pl ant S yst em Des cr i ption s, IB-4, p. 6 ANSWER 2 12 (1 00) answera b. (1.0)
REF ER E NC E VCS P lant System Descriptions, 18-4, p. 7 ANSWER 2.13 (1 00) a ns we r s c. (1.0)
R EF ER EN CE VCS Plant S ystem Des cr iptions, IB-4s pp 15, 17 ANSWER 2 14 (1.00) answers c. (1.0)
REFERENCE VCS Pl ant Syst em Des cr lptions, IB-4, p. 21 A NS WER 2.15 (1.50) answera es d, g [0.5 each] (1 5)
R EF ER ENC E VCS Plant System Descriptions, IB-5s p. 33
r Za._EL&MI_QE11GM_IMCLUDIHE SAEEIX_ASQ_EBERGEMC1.1111E51 PAGE 44 ANSWERS -- SUMMER -8 5 / 02/ 2 5-P IC KE R, 8.
ANSWER 2 16 (1.00) answers c. (1 0)
REFERENCE .
VCS Pl ant Syst em Descr ip tion s, IB-5s p. 32 ANSWER 2.17 (1 00) answer 8 b. (1.01 REFERENCE VCS P la nt System De sc rip tions, 18-5, p. 49 A NSWER 2.18 (1 00) a ns wa r s b. (1.0)
R EF ER EN CE VCS Pl ant S yst em Des cr iptions, GS-2s p. 22 ANSWER 2 19 (1.00) answers a. (1.0)
RE FE RE NC6-VCS Pl ant Syst em Descr ip tion s, GS-2, p. 28 A NS W ER 2 20 (1.00) answer b. (1 0)
R EF ER ENC E VC S P l ant Sy stem Des cr ip tion s, GS-2, p. 19
~ 24__EL&MI_QEllGN_1HCLUDING_1&EEIX_&HQ.E5ERGE1CL_1111E51 PAGE 45 ANSWERS -- SUMME R -85/ 02/ 25-P ICKE R, B.:
ANSWER .2.21 (1.00) answer k c, g g d M (1.0)
REFERENCE -
VCS Pl ant Syst em Descr iptionss TB-7, p. 22 ANSWER 2.22 ( 1. 00 )
answers c. (1.0)
REFERENCE VCS P l ant Sy stem De s c r i p t i on s, TB-7, p. 24 A NSW ER 2.23 (1 00)
Ons we r s a. 3
- b. I
- c. 2 [0.33 each) (1 0)
REFERENCE VCS Pl ant S yst em Descr iptionss 05-8, p. 9
I
- 1s..I il&MBEhli_AdQ CQUIgQL1 PAGE 46 ANS WERS -- S UMM ER -85 /02/25-PICK ER, B.
l l
1
- ANSWER 3.01 (1.00) a ns we r s a. (1 0)
R EF ER ENCE -
VCS Pl ant System Des cr iptions, I C - 1, p . 17 ANSWER 3 . 02 (2.00)
- a. 3
- b. 4
- c. 2
- d. 1 C0.5 each) (2.0)
REFERENCE VCS Plant. System Descriptions, IC-1, pp 19-23 ANSWER 3.03 (1 00) answers b. (1 01 REFERENCE VCS Plant System Descriptions, IC-2, p. 10 ANSWER' 3.04 (1.00) answers c. (1.0)
REF ERENCE VCS Plant System Descriptions, IC-2, p. 23 ANSWER 3.05- (1.00) answers a.' (1 0)
REFERENCE VCS Plant-System Descriptions, IC-2, pp 24-28
~1s _INSIEMBENIl_&BR_CQHIEQLS PAGE 47 ANSWERS -- SUMMER -35/02/25-PICKERS B.
ANSWER 3.06 (1.00) ans we r s b. (1.0)
REFERENCE -
VCS Plant System Descriptions, IC-3s p. 9 ER 3.07 (1.00) answers . (1.0)
REFERENCE VCS Plant System Descriptions, IC-3, p. 34 NANSW 3.08 (1 0 )
answera c. O E (1.0)
REFERENCE
'/CS. Plant System Descriptions, IC-3s-pp 33-35s Fig. IC 3.13 A NS W S.R' 3.09 (1.00) ans wer s ' b. (1.0)
REFERENCE
'VCS Plant System Descriptions, IC-3, pp 35-40 ANSWER 3 10 (1.00) answert b. (1 0)
REFERENCE VCS Plant System Descriptionss IC-9s p. 47
'la__lH11RudENII_AHQ.CQHIRQL1 PAGE 48 ANS WERS -- SUMMER -85/02/25-PICKER, B.
ANSWER 3.11 (1.00) l answers c. (1.0) l REFERENCE ".
VCS Plant System Descriptions, IC-6, p. 25
. ANSWER 3 12 (2.00) answers a. FALSE *
- b. TRUE
- c. TRUE
- d. FALSE CO.5 each] (2 0)
REFERENCE VCS Plant System Descriptions, IC-8, pp 14, 30, 39, 42
-ANSWER 3 13 (1.00) answers b. (1.0)
REFERENCE VCS Plant System Descriptions, IC-9, pp 13-15 ANSWER 3.14 (3.00) answers a. 3
- b. 4
- c. 5
- d. 8
- e. 2
- f. 1 ;_ [0.F each) (3.0)
REFERENCE VCS Plant System Descriptions, IC-9, pp 44-46, 82-85 ANSWER 3.15 (1.00) answers c. (1 0)
35._181ItuBEK11_AND_CDHIRQ13 PAGE 49 ANSWERS -- SUMMER -85/02/25-PICKER, B.
' REFERENCE VCS Pl ant System Descriptions, IC-9s pp 58-59 ANSWER 3.16 (1.00)
. answers c. (1 0) '.
' REFERENCE VCS Plant System Descriptions, IC-9s pp 25, 26 4
ANSWER 3 17 (1.00) answers a. (1 0)
R EF ER ENCE
-VCS Plant System Descriptions, IC-12, pp 4-10 ANSWER 3 18 (1 00) ans we r s d. (1 0)
REFERENCE VCS Plant Syst em Descr iptions, IC-13, p. 4 1
- ANSWER ' 3.19 (1.001 answers a.' (1 0)
REFERENCE l VCS Plant System Descrip tions, IC-13, p. 15
' ANSWER '3.20 (1 00) l answer Kd, (1 0)
REFERENCE F1 VCS Pl ant . System Descr iptionss IC-2s Review Questions and Key
"la- 1811EW5EH11_AND_CQUIEDL3 PAGE 50 ANSWERS -- SUMMER -85/02/25-PICKER, 8.
ANSWER -3.21 (1.00) answers c. (1,o)
' REFERENCE .
VCS Plant System Descriptionss IC-9s p. 58
"sa__EEQCEQuRE1_=_BDE5Att_AADDEd&La_EBERGENCE_AND PAGE 51
&&Q10LQGICAL_EDNIEDL ANS WE RS -- SUMME R -85 / 02/ 2 5- PIC KE R, B.
ANSWER 4.01 ( 1. 00) ans wer s b. (1.0)
REFERENCE VCS SOP-101s p. 1 ANSWER 4.02 (1 00) anwcra a. (1.0)
R EF ER ENCE VCS SOP-102, p. 1 ANSWER 4.03 (1.00) an s we r s c. (1.0)
REFERENCE V CS S OP-102s p. 1 ANSWER 4.04 (1.00) answers c. (1 0)
RE F ERE NCE VCS GOP A ppend ix As p. 4 Plant System Descriptions, IC-5, number of Shutdown b anks.
A NS WER 4.05 (1.00) ans we r s b. (1.0)
REFERENCE i
---e b - --
- i s__ ERQCEDU E El_=_ bD85& L A_ AabO & d&L a_EEEEE ESC 1_ AS D PAGE 52
.RADIDLDEICAL_CDNIRQL ANSWERS -- SUMMER -85/02/25-PICKER, B.
ANSWER 4.06 (1.00) answers a. (1.0)
REFERENCE VCS E0P-1.3, p. 1 ANSWER 4.07 (1.00) ans wer s b. (1.0)
REFERENCE VCS E0P-2.0, p. 3 ANSWER 4.08 (1.50) answers h 1,6 4 Cr O 'N ' ' "" - ~
(1.5)
REFERENCE VCS E0P-1.2, pp 4, 5 ANSWER 4.09 (1.00) answers a. (1.0) l i
I REFERENCE VCS GOP Appendix As p. 17 ANSWER 4.10 (1 00) answers d. (1.01 REFERENCE VCS GOP-1, App en di x As p. 4 l
1 l
l
r ia__ERDGEQUEE1_=_HDR5&La_ABBQEd&LA_EHEREENCI.ANQ PAGE 53 RA21DLDEICAL_CDMIEDL ANSWERS -- SUMMER -85/02/25-PICKER, B.
ANSWER 4.11 (1.00) cnswert d. (1.0)
REFERENCE VCS E0P-1.Os p. 3 ANSWER 4.12 (1.00) answert a. (1.0)
REFERENCE VCS GOP Appendix C, p. 1 ANSWER 4.13 (1 00) answers c. (1 0)
REFERENCE VCS GOP-3, p. los note 10 ANSWER '4.14 (2.50) answert a. 2
- b. 3
- c. 1
- d. 1
- e. 6 [0. 5 e ach) (2.5)
REFERENCE VCS Radiation Protection Fundamentals, Section 5, pp 5-12 ANSWER 4.15 (1.00) answers c. (1.0)
r eks _REQGEQURES.=_HQRHALa_AAEDREALa_EBERGENC1_AMQ PAGE 54 RAQ10 LOGICAL _EQ5IRQL ANSWERS -- SUMMER- -85/02/25-PICKER, B.
REFERENCE.
. ANSWER 4.16 (1 00) ansters- c. (1.0)
-REFERENCE VCS Radiation Protection Fundamentalss Section los pp 44-45 ANSWER 4.17 (3.00) answers a. 3
- b. 5
- c. 2-
- d. 1
- e. 6
- f. 4 [0.5 each) (3.0)
REFERENCE VCS' SOP 102- s 210-33, 306-52, 401-35, 57 AOP 11.0-1 E0P 1.0 ANSWER 4.18 (2.00) answers a. FALSE
- b. FALSE
- c. TRUE
- d. TRUE Co.5 each) (2.0)
REFERENCE VCS E0P-10.0, pp 12-15 ANSWER 4.19 (1.00) answers b., (1.0)
REFERENCE VCS E0P-18.2, pp 9, 11
o dia. EEQCEDURE1_=_dQB5&La_ABBORMALa_EBEREENC1_&ME PAGE 55
? EAD10 LOGICAL.CQHIRQL ANSWERS -- SUMMER -85/02/25-PICKER, B.
ANSWER 4.20 (1.00) ansters Subcriticality ( 3)
Core cooling (5)
- Heat Sink (2) .
Integrity (6)
Containment (1)
Inventory (4) (1.0)
REFERENCE ~
VCS EDP-12.0, p. 1
< UNITED STATES 4 [p ErgS.o NUCLEAR REGULATCRY COMMISSICN
! f)
- CE! ION 11 ,
101 MARIETTA STREET,N.W. j
'E-
- 0 In [ ATLANTA, GEORGI A 30323 l
\...../ U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION Facility: V. C. Summer Reactor Type: Westinghouse, 3 Loop D1te Administered: February 25, 1985 Examiners: T. Rogers '
Applicant: -
INSTRUCTIONS TO APPLICANT:
Use~ separate paper for the answers. Write answers on one side only. Staple question sheets 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.
Category Value Total Score Value Category 25% 32 5. Theory of Nuclear Power Plant Operation, Fluids and Thermodynamics 25% 6. Plant Systems: Design, Control &
Instrumentation 25% 32 7. Procedures-Normal, Abnormal, Emergency Radiological Control 25% 32 8. Administrative Procedures, Conditions and Limitations 100% 128 TOTALS Final Grade %
All work done on this exam is my own, I have neither given or received aid.
Applicant Signature e e I
O-
\ -:
- o. .f b.
a
+ b 5.0 Theory of Nuclear Power Plant Operation, Fluids and Thermodynamics (32.0) 5.1~ The units of neutron flux is (1.0)
(a) neutrons 4 - sec -
o (b) neutron-cm
, cm3-sec (c) neutrons , ,
Cm2 (d) . neutrons em- sec 5.2 During fuel loading, which of the following will have (1.0) no effect on the shape of a 1/M plot?
I (a) The location of the neutron sources in the core.
(b) The strength of the neutron sources in the core.
l-l (c) The location of the neutron detectors around the core.
(d) The order of placement of fuel assemblies provided the proper enrichments are placed in their proper location.
l 5.3 If reactor power increases from 1000 cps to 5000 cps in 30 (1.0) seconds, what is the SUR?
l.
~ '
0 2
l
)
5.4 An ECP is calculated for a reactor startup 4 hcurs after (1.0) a reactor trip from 100% equilibrium condition. Which one of the following conditions would cause the actual critical position to be lower than the ECP?
(a) The startup is delayed until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the trip.
(b) Actual boron concentration is 10 ppm more than the predicted boron concentration. ,
(c) A rod finger is separated from its spider assembly (d) The steam dump pressure setpoint is lowered by 100 psi prior to reactor startup.
5.5 The change in reactivity associated with a change (1.0) in Keff from 0.920 to 1.004 is approximately (a) 0.091 (b) 0.084 (c) 0.087 (d) 0.080 5.6 Which of the following is NOT a true statement (1.0) concerning Xenon poisoning?
(a) The concentration will buildup and insert negative reactivity immediately following a reactor trip.
(b) The time after trip that Xenon peaks is independent of neutron flux before the trip.
(c) Equilibrium Xenon reactivity worth at 50%
power is NOT half of the equilibrium Xenon
- reactivity worth at 100% power.
(d) 100% power Equilibrium Xenon reacti'vity worth is higher than 100% power equilibrium samarium Feactivity worth throughout core life.
3 4,
5.7 Duiirg power operation, with rods in Manual, a positive
~
(1.0) peak in the AFD caused by Xenon can be suppressed by (a) Running with a reduced Tavg.
~(b) Partial insertion of control Bank D.
(c) Oscillating. reactor power. ,
(d) Baron dilution.
-5.8 .The equilibrium subcritical multiplication level of a reactor (1.0) i with a source level of 100 neutrons and a keff of 0.95 is: l sec l
(a) 2.1 x 103 counts /sec (b)- 2.05 x 102 counts /sec (c) 1.05 x 102 counts /sec (d); 2 x 103 counts /sec
! 15 . 9 Concerning the behavior of Samarium-149, which one -(1.0) of the following statements is true?
(a) Once equilibrium Samarium is established, Samarium reactivity worth does not chance regardless of power level changes.
L l- (b) 50% equilibrium Samarium reactivity worth is equal
( to 100% equilibrium Samarium reactivity worth.
l- (c) Samarium is only removed by radioactive decay.
(d) Samarium is produced by the decay-of Iodine.
u I.
o I- -
4
.5.10 The reactor is operating at 50% power with the rod (1.0) control system in manual when a single Group A rod drops into the core. . Assuming no reactor trip or ,
operator actions occur, choose the answer that best {
describes the final steady state conditions.
(a) Final power = initial power, final Tave less than initial Tave.
(b) Final power = initial po..er, final Tave = initial -
Tave.
(c) Final power less than initial power, final Tave greater than initial Tave.
(d) Final power = initial power, final Tave greater than initial Tave.
5.11 Which of the following is a true statement concerning (1.0) control rods?
l (a) The rod reactivity worth is highest at its axial centerline.
1 (b) Two adjacent rods will reduce each others reactivity worth.
(c) Rod reactivity worth is independent of boron con-centration, at a constant position.
(d) The rod reactivity worth is the primary control L to compensate for fuel burn-up.
I
!- 5.12 Which'of the following factors of the six factor formula is (1.0)
- . greater than 1.0 for V. C.. Summer Nuclear Plant?
(a) Fast fission factor.
(b) Resonance escape probability.
(c) Thermal non-leakage probability.
t
-(d) Thermal utilization factor.
, + 5' 5.13 The effective delayed neutron fraction decreases over core life (1.0) partly because:
(a) the number of delayed neutron precursor groups increase.
239 (b) the fission yield for Pu -
increases.
239 (c) of a buildup of Pu in the core.
(d) soluble boron is removed from the core.
5.14 Which of the following will NOT change over core. life? (1.0)
(a) The acceptable AFD target band as a function of power.
(b) The minimum acceptable shutdown margin.
(c) The control rod reactivity worth.
(d) T'he power defect reactivity worth.
5.15 Which of the following is a true statement concerning' (1.0) the moderator temperature coefficient?
(a) The MTC tends to drive the neutron flux toward the top of the core over core life.
(b) The MTC increases (more negative) as the-boron concentration increases.
(c) The MTC effects the axial neutron flux distribution more than the radial neutron flux distribution.
(d) The MTC will not change with a change in temperature if the boron concentration is maintained constant.
5.16 Which of the .following statements is true concerning soluble (1.0) boron reactivity control?
(a) As boron concentration increases, the differential boron worth (PCM/pom) decreases.
(b) As Tavg increases, boron concentration (ppm boron) decreases.
(c) As Tavg increases, the differentia 1' boron worth (PCM/ ppm) increases.
(d) As fission products buildup, differential boron worth (PCM/ ppm) increases.
b- -.=..i_
6 5.17 Which of the following is a true statement concerning (1.0) radioactive decay? Remember the atomic number is the number of protons and the mass number is the number of neutrons plus protons.
-(a) When an element decays by beta emission, the new element will have increased in atomic number by one and the mass number will remain the same as the original element.
(b) When an element decays by alpha emission, the new -
element will have decreased in atomic number and .
mass number by two, from the original element.
(c) When an element decays by neutron emission, the new element will have increased in atomic number by one and decreased in mass number by one, from the original element.
(d) When an element decays by gamma emission, the new i element will have increased in atomic number by one and the mass number will remain the same as the original element.
5.18 Which of the following is NOT a criterion for emergency core (1.0) cooling systems as specified in the Federal Regulations?
(a) Limiting the maximum cladding oxidation to less than 17%.
(b) Limiting the peak clad temperature to less than 2200 F.
(c) Ensuring a coolable core geometry is maintained.
(d) Limiting the maximum hydrogen generation to less than 50% by volume in containment.
5.19 The largest contribution of hydrogen released to containment (1.0) during an accident involving inadequate core cooling and reactor vessel void formation is from (a) a zirconium - steam reaction.
(b) an aluminum - steam reaction.
(c) the release of dissolved hydrogen in the coolant from the hydrogen overpressure on the volume control tank.
(d) radiolysis of the coolant.
7 5.20 One reason for the requirement of a minimum control rod worth is: (1.0)
(a) to minimize the adverse effects on DNBR from a dropped rod.
(b) to minimize the adverse effects on core power from a stuck
' rod upon a reactor trip.
(c) to minimize the adverse effects on core power upon a rod ejection accident. ,
(d) to overcome the power defect upon a reactor trip.
5.21 A variable speed centrifugal pump is operating at 1800 rpm (1.0) with a capacity of 400 gpm at a discharge head of 20 psi which requires a power of 40 kW. If pump speed is increased to 2000 rpm, which of the following best describes the new pump parameters?
(a) 444 gpm, 22 psi, 55 kW (b) 444 gpm, 25 psi, 49 kW (c) 444 gpm, 25 psi, 55 kW (d)--444 gpm, 22 psi, 49.kW 5.22 Which of the following is true of pump-operation? (1.0)
(a) A' centrifugal pump motor will draw more power if it is started with the discharge valve shut, than it would if the discharge valve is open.
(b) Unless specified, a positive displacement pump should be started with its discharge valve shut.
(c) Only centrifugal pumps have a required net positive suction head.
(d) The flow through a positive displacement pump will not significantly change with varying discharge pressure.
1 1
l
9-I 8
l 5.23 The condensate subcooling in a condenser operating (1.0) l l
at 1 psia with a condensate temperature of 95'F !
is approximately:
(a) 6.7*F (b) 196.7'F l (c) 1.07*F .
-(d) 25.3*F .
5.24 Which of the following is true of the bases'for the Tech Spec (1.0)
Presssure/ Temperature limits.
(a) During cooldown, the most limiting location is at the outside of the vessel wall because of the thermal gradients produce tensile stresses there.
(b) During cooldown, the most limiting location is at the inside of the vessel wall because the thermal gradients produce tensile stresses there.
(c) During heatup, the most limiting location is at the inside of the vessel wall because the thermal gradients produce compressive stresses there.
(d) During heatup, the most limiting location is at the outside of the vessel wall because the thermal gradients produce compressive stresses there.
5.25 Which of the following is NOT true concerning (1.0) heat exchangers?
(a) Heat transfer is by both the conductive and convective methods of heat transfer.
(b) The heat transfer rate for a parallel flow heat exchanger is higher than that of a counter flow heat exchanger under the same inlet temperatures.
(c) The heat transfer rate is directly proportional to the heat transfer coefficient associated with material the tubes are made of.
(d) Higher thermal stresses across the tubes will ;
accompany higher tube thickness.
p s
9 5.26 Which of the following boiling regimes is acceptable (1.0) during normal reactor operations?
'(a) Transition boiling (b) Film boiling (c) Partial film boiling (d) Nucleate boiling 5.27 The need to change the RTNDT of the reactor vessel over (1.0) the life of the plant is-a result of:
(a) thermal cycles (heatup and cooldown transients).
(b) pressure cycles (changes in pressure).
(c) gamma irradiation.
(d) neutron irradiation.
5.28 The rod bow penalty used in calculating the nuclear (1.0) enthalpy rise hot channel factor-is a function of:
(a) total core flow.
(b) fuel burnup.
(c) nuclear power.
(d) reactor coolant system pressure.
5.29 Which of the following is a true statement if the (1.0) power range channels were adjusted based on a calculated calorimetric with the given errors.
(a) If the feedwater temperature used in the calorimetric calculation was lower than actual feedwater. temperature, actual power will be higher than indicated power.
(b) If the reactor coolant pump heat input used in the calorimetric calculation was neglected, actual power will be' greater than indicated power.
(c) If the steam flow used in the calorimetric calculation was lower than actual steam flow, actual power will be less than indicated power.
(d)_ If one of the steam generators were omitted from the blowdown heat rate calculation, the actual reactor power will be greater than indicated power.
9 v-4 .
r:
10 5.30 During a plant heatup with the pressurizer pressure at (1.0) 1000 psig, failure of the air supply solenoids allows the PZR PORV to open slightly to a throttling position. The maximum pressure reached downstream of the valve is approximately the PRT pressure of 5 psig. What would be.the condition of the fluid downstream of the valve?
(a) Superheated steam (b) Saturated steam (c) Wet vapor
'(d) Subcooled liquid 5.31 Which of'the following products discribes the heat transfer (1.0)
-rate across the' steam generttor tubes?
(a) mcpAT
.(b) UAAs (c) mcp.Ah (d) UAAT Where m is the. mass cp is the specific heat AT is the change in temperature Ah is the change in enthalpy As is the change in entropy U is the heat transfer coefficient 5.32 Specific entropy of steam entering the low pressure turbine (1.0)
(a) will always be higher than the specific entropy of steam exiting the LP turbine.
(b) will always be lower than the soecific entropy of steam exiting the LP turbine.
(c) will be higher or equal to the specific entropy of steam leaving the.LP turbine. .
(d) may be higher than, lower than, or equal to the entropy of steam exiting the LP turbine.
Write "End of Section 5.0" on your answer sheet.
2 .h
11 6.0 Plant Systems Design, Control,-and Instrumentation (32.0) 6.1 The steam generator PORVs will (1.0)
(a)' respond as a bank 4 steam dump when the control' transfer switcher are in the PWR RLF position.
(b) _ fail.to the steam dump mode on loss of selector control signal.
~
_(c) fail open on loss of control air.
(d) respond to commands of the pressure controller and the control board AUT0/ MANUAL station when the control transfer switches are in the AUTO position.
6.2 The input to.the feed pump speed controller programmed AP is- (1.0)
_(a) Pimp (b) Total core AT (c) Total compensated steam flow.
(d) Auctioneered high Tavg.
6.3 Which of the-following statements concerning the (1.0)
Containment Spray System'is true?
(a) The sodium hydroxide in the spray facilitates the converting of_ soluble iodine into insoluble iodine.
(b) Upon reaching the low-low level setpoint in the RWST, the suction of the spray pumps will be automatically shifted to the contai~nment sump and isolated from the RWST.
(c) During the injection phase of spray operation, the spray additive orfices limit the sodium hydroxide in the spray.
to a maximum pH of 10.5 to limit corrosion inside of containment.
(d) The spray' pumps start on a SI signal, but they do not inject into containment until the discharge valves are opened upon receipt of a high reactor building internal pressure on 2/4 channels.
4
- - 12 6.4 Wh'ich'of the following statements concerning the (1.0) pressurizer safety valves is NOT true?
(a). A water seal will normally be maintained below each valve seat to reduce the problem of valve seat leakage development.
(b) Each valve has a design capacity greater than the maximum surge rate resulting from a complete loss ,
of load without a direct reactor trip.
(c) The three safety valves discharge to a common manifold which discharges into the PRT.
(d) -Each safety valve has a temperature indicator in its discharge line to indicate the passage of steam due to leakage or valve lifting.
' 6.5 Concerning the letdown isolation and orifice isolation (1.0) valves, which of the following statements is true?
(a) The. letdown isolation valves can be closed from the MCB with the orifice isolation valves open.
(b) The letdown isolation valves automatically open if the pressurizers' water level is greater than 17% and both letdown isolation valves are shut, when the "0 PEN" position is selected.
(c) These valves fail-as-is on a loss of power.
(d) The letdown isolation valves must be open before opening an orifice isolation valve from the MCB.
6.6 Which of the following sets of pressurizer pressure (1,0) setpoints is correct?
(a) 2335 psig - PORV opens 2385 psig - High Pressure alarm 2435 psig - High Pressure trip (b) 2210 psig - B/U heaters on 1865 psig - Low Pressure alarm 1850 psig - Low Pressure trip
-(c) 2380 psig - High Prer 're trip 1870 psig - Low Prer;u.e trip 1850 psig - Low Pressure SI (d) 2335 psig - PORV opens 2100 psig - PORV blocked 2000 psig - Low Pressure alarm
13 6.7 The pressurizer pressure and level control system is (1.0)-
designed to accommodate without a trip.
- (a) loading or unloading at 10% per minute (b) _ instantaneous load changes at 120 percent without automatic rod control (c) a step load' reduction of 95% with automatic rod control and steam dump operation -
-(d) loading at 10% per minute and unloading at 15% per minute 6.8 The reactor trip breaker shunt coils will: (1.0)
(h)energizeonanyautomatictripsignal.
(b) _ energize on only SI initiated automatic trip signals and manual trips.
(c) energize only on manual trip signals.
(c) energize on only RPS and SI initiated automatic trip signals.
- 6.9 The movable incore fission chamber detectors operate in the (1.0)
(a) recombination region.
(b) ionization region. -
(c) proportional region.
(d) GM region.
6.10 The Moisture Separator Reheater's reheating steam is supplied (1.0) by:
(a) Main steam.
(b) Auxiliary steam.
(c) HP turbine extraction steam.
(d) HP turbine exhaust steam.
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14 6.11 The Cir:ulating Water System open cycle booster pumps do NOT (1.0) supply cooling water to the:
(2) vacuum pump coolers.
(b) turbine oil coolers.
(c) hydrogen coolers.
(d) auxiliary condensers. ,
6.12 Which of the following is true concerning the Source . (1.0)
Range channel high voltage cutoff? .
(a) During a reactor startup either IR channel increasing above P-6 will turn off the high voltage.
(b) If one IR channel fails low while at power, the Source Range high voltage will be re-energized.
(c) Two out of four PR channels above 10% power will block the high voltage.
(d) During a reactor shutdown 11ther IR channel decreasing below P-6 will turn on the high voltage.
6.13 Which of the following statements about the Digital (1.0)
RcJ Position Indication (DRPI) System is correct?
(a) The DRPI system uses a series of magnetic switches to determine rod position.
(b) Anytime rods within a group differ by more than 12 steps a ROD DEVIATION alarm'is generated.
(c) A control bank B rod on the bottom will not generate a RPI R00 AT BOTTOM alarm unless control bank B rods are at least 12 steps.
(d) Power to the DRPI system is supplied from the control rod MG sets.
15
- 6.14 The one input to the rod insertion limit calculator (1.0) i i
used as an indication of reactor power is:
(a) Pimp k
(b) Auctioneered High Tavg (c) Total Steam Flow (d) Auctioneered High Delta T ,
6.15 A major difference between an ion chamber and a G-M .
(1.0) detector is:
(a) the G-M detector has a photo-multiplier tube to increase it's sensitivity.
(b) the ion chamber is filled with a gas and the G-M detector operates under a vacuum.
I (c) the ion chamber operates at such a low voltage that a significant number of ion pairs are lost by recombination, thereby decreasing its sensitivity.
(d) the G-M detector operates at a much higher voltage causing gas multiplication to increase the charge collected to a value independent of the ionization initiating it.
6.16 Which of the following is true concerning the turbine generator (1.0) electrohydraulic control speed control unit?
(a) The wobulator input is in effect only when 1500 rpm is selected as the speed reference.
(b) The loss of 1 out of 2 speed signals will trip the turbine.
(c) The Medium STARTUP RATE of 5%/ minute is automatically selected when the turbine is reset.
(d) The secondary acceleration amplifier and integrator provide inputs to the control circuit only when turbine speed is less than 100 RPM.
6.17 What type of radiation does the Primary Coolant Letdown (1.0)
-Monitor detect for indication of fuel failure?
(a) Alphas
-(b) Betas (c) Gammas ,
(d) Neutrons
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16 6.18 Which of the following statements concerning the S/G (1.0) blowdown Processing System is correct?
(a) A radiation monitor is located in each S/G blow-down line.
(b) S/G blowdown monitor (RM-L3) will automatically divert blowdown flow to the nuclear blowdown holdup
~
tank upon high activity alarm.
(c) Tte blowdown surge tank is bypassed unless the blow- -
down radiation level is greater than discharge limits.
(d) A S/G blowdown high radiation level on 'SG Blowdown discharge monitor (RM-LIO) will automatically shift the blowdown discharge from the circulating water system to the main condenser.
6.19 Which of the following is NOT a function of the P-4 (1.0) permissive (trip and bypass breakers open)?
(a) Allows bypassing of steam dump cooldown interlock.
(b) Blocks SI after actuation and SI reset.
'(c) Causes feedwater isolation if a low Tavg signal is also present.
(d) Causes a turbine trip.
6.20 Which of the following conditions should generate a safety JW2E11 (1.0) injection signal?
(a) Loop A steam line pressure at 670 psig with loops B and C steam line pressure at 700 psig.
(b) Pressurizer level less than 15%.
(c) Reactor coolant loops A and B at 550 F and Loop C at 553 F.
-(d) Reactor coolant loops A and B with 80% flow and loop C with 100% flow.
17 6.21 The RHRS hotleg isolation valves will: (1.0)
(a) allow manual opening if Pressurizer pressure is less than 425 psig.
(b) automatically open if the RCS pressure at the hot leg is less than 425 psig.
(c) allow manual opening if the RCS pressure at the hot leg is less than 425 psig. ,
(d) automatically open if pressurizer pressure is .
less than 425 psig.
6.22 Which of the following valves will automatically (1.0) open by a safety injection signal?
(a) Boron injection tank recirculation isolation valves, if shut.
(b) RHR pump motor operated discharge valves, if shut.
(c) RWST to RHR pump isolation valves, if shut.
(d) Accumulator discharge valves, if shut and power available.
6.23 The hotwell level is controlled by: (1.0)
(a) self cavitating hotwell pumps responding to available NPSH.
(b) variable speed hotwell pumps responding to the hotwell level signal.
(c) supply and return valves between the hotwell and condensate storage tank responding to the hotwell level signal.
(d) balancing feed flow and steam flow.
6.24 The turbine-driven EFW pump will automatically start on: (1.0).
(a) a safety injection signal (b) low-low steam generator water level in two steam generators (c) loss of three main feedwater pumps (d) loss of power to one emergency bus.
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18 6.25 Which of the following is a rod control interlock? (1.0)
(a) The C-3 interlock will stop rod withdrawal in automatic but not in manual when the OTAT signal is within 3% of the calculated trip value.
(b) The C-2 overpower rod stop blocks automatic and manual control rod withdrawal when one power range channel exceed 103%. .
(c) The-C-5 interlock ensures that the rod control .
system is not placed in automatic until 10%
turbine power is attained.
(d) The C-4 interlock will stop all rod motion in automatic or manual when the OPAT signal is within 3% of the calculated trip value.
6.26 Starting the diesel. generator using the EMERGENCY START (1.0) push botton will block any diesel generator trip signal.
(a) engine overspeed (b) low lube cil pressure (c) high lube oil temperature (d) generator differential overcurrent 6.27 Which of the fo?'owing will be started by the ESF loading (1.0) sequencer for an SI signal but NOT a blackout signal?
(a) RHR pumps (b) Service water pumps (c) Component cooling water pumps (d) Motor driven emergency feed pumps
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19 6.28 Two separate sources of off-site power are provided for the (1.0) safeguards power electric 5.> stem. One is by the SCE&G trans-mission grid and the other is by (a) the Fairfield Pumped Storage Facility.
(b) Pineland-Station.
(c) S.C.P.S.A. Blythewood Station. ,
(d) Parr Generating Complex.
'6.29 The gaseous waste processing system is comprised (1.0) of:
(a) three waste gas compressors, eight gas decay tanks, and three recombiners.
(b) two waste gas compressors, seven gas decay tanks, and three recombiners.
(c) three waste gas compressors, seven gas decay tanks, and two recombiners (d) two waste gas compressors, eight gas decay tanks, and two recombiners.
6.30 Prior to a radioactive liquid release, the liquid (1.0) is circulated using a in order to obtain a representative sample before discharge.
(a) Waste evaporator feed pump (b) Waste evaporator condensate pump (c) Floor drain tank pump (d) Waste monitor tank pump 6.31 Evaporation losses from the spent fuel pit is made up from (1.0)
(a) the refueling water storage tank.
(b) the volume control tank.
(c) the demineralized water system.
(d) the boric acid tank.
Section 6.0-is continued on the next page
20 6.32 Which of the following refueling components is (1.0) peumatically operated?
(a) Thimble plug handling tool latching mechanism.
(b) The opending machine lifting arms.
(c) Pusher arm drive motor.
(d) Control rod change fixture gripper. '
Write "END OF SECTION 6.0" on your answer sheet.
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21 7.0 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL (32.0)
CONTROL 7.1 Which of the following is NOT a requirement for opening the RCP (1:0) seal water bypass valve during RCS pressurization?
(a) The RCS is greater than 100 psig, but less-than 1000 psig.
(b) .The No. I seal leakoff valve is open.
(c) The No. I seal leakoff flow rate is less than 1 gpm.
(d) The RCP is in operation.
7.2. Which of the following is the maximum allowed primary to steam (1.0) generator differential pressure?
(a) 1400 psid (b) 1600 psid (c) 1800 psid (d) 2000 psid 7.3 Which of the following statements concerning the use of Inverse (1.0)
Count Rate Ratio (ICRR) plots is correct?
(a) ICRR plots are required for all startups where the reactor has been shutdown for less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(b) The ICRR value is calculated by dividing 1 by the observed count rate.
(c) Rod withdrawal increments between successive ICRR data points should not'be more than 50 steps.
f (d) Count rate data must be taken off the NI counter-scaler with a minimum count time of one minute.
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22 7.4 -After criticality is established during a reactor startup, the (1.0) 8 reactor power is increased to 10 amps for critical data by (a) dumping steam to the condenser.
(b) placing a feed pump in service.
(c) withdrawal of control rods.
(d) boron dilution. .
7.5 Which of the following is true concerning a reactor startup? (1.0)
(a) A sustained startup rate of 0.5 decades per minute should not be exceeded.
(b) Shutdown rods must be verified to be fully withdrawn within
- 15 minutes prior to the anticipated control rod bank withdrawal.
(c) The ECP must be recalculated if the startup is delayed one hour after the anticipated time of criticality.
(d) Prior to paralleling the generator to the grid, nuclear power is increased to 15% by dumping steam to the condenser with steam dumps in the Tavg Mode.
7.6 Which of the following is true concerning RHR system operations? (1.0)
(a) During RHR system startup, a warmup period is required to limit thermal shock to components.
(b) The boron concentration in the RHR system should be less than the RCS boron concentration before placing the RHR system in operation.
(c) The cooldown rate is controlled by varying the CCW flow to the RHR heat exchanger.
(d) The RHR System may be valved into service at 400 psig, but recirculation cannot be established until RCS pressure is less than 175 psig because of the RHR pump designed shutoff head.
b
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23 7.7~ In accordance with GOP-1, a steam' bubble being formed- .(1.0) in the pressurizer is indicated when (a) the PRT. level is' constant and the PRT pressure is increasing.
(b) the-PRT pressure is constant and the PRT level and
' temperature is increasing.
(c) charging flow is greater than letdown flow.with RCS '
pressure: constant.
.(d) charging flow is less than letdown flow with RCS pressure constant.
L7.8 When performing a startup of a main feedwater pump, the turbine .
_( 1.0) speed is , determined to have matched the feedwater fspeed
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obsuvat
.(a) The' FEEDWATER PUMP / SPEED CONTROLLER MISMATCH error indicator.
.(b) feed flow balancing with steam flow.
(c) a constant SG level.
(d) the applicable feedwater pump speed.
7.9 Which of the following statements concerning.the Immediate Actions- (1.0)
-for nuclear instrument malfunctions is correct?.
(a) If a source range channel fails below P-6, increase power to above P-6 if a startup is in progress.
.(b) If an intermediate range channel fails above P-6 but below P-10, reduce power.to below P-6.
(c) If the audio count rate channel fails during a reactor star. tup, the control banks must be inserted to O steps.
(d) If both source range channels fail while below P-6, borate the RCS to the shutdown margin requirements of the Technical Specifications, i-t_.
24
- 7.10 Which.of the'following statements about the Dropped Rod procedure (1.0) is correct?
.(a) The P to A converter readino must tme recorded from the indication on the main contral board.
(b) .The step counter for the affected bank must be manually reset to zero prior to an attempt to withdraw the dropped rod.
(c) Upon starting recovery of the dropped rod, a NON URGENT -
FAILURE alarm will-occur because the lift coils for the other rods in the group have been disconnected.
(d) If in manual rod control, place turbine DEH control in HOLD.
7.11 Which of the following is NOT true when barring the diesel (1.0) generator.
(a) The engine'must be barred at least one complete revolution for proper checkout.
(b) The engine may be barred using a wrench attached to the-shaft end.
(c) The engine may be barred using the barring device motor.
(d) The engine may be barred using a starting Air-Install lever on a start air control valve.
7.12 Which of the following describes the procedure for calculating (1.0) a quadrant power tilt ratio?
(a) Larger of Maximum Iu or Maximum IL Minimum Iu Minimum IL (b) Larger of Maximum Iu or Maximum IL Average Iu Average IL (c) Larger of Maximum Iu - Minimum Iu Maximum IL - Minimum IL' or (d) The Largest Iu Calculated for each operable power IL range channel.
Where I represents the normalized detector reading u represents upper _ detector reading L represents lower detector reading
25 7.13 Which of the following is a required operator action prior (1.0) to exiting the main control room in accordance with the control room evacuation procedure?
(a) . Implement the Emergency Plan, Alert Conditions.
(b) Emergency borate.
(c) Start the EFW pumps. .,
(d) Trip the main feedwater pumps.
7.14 If a reactor coolant pump trips at 20% power, the operator should. (1.0)-
(a) determine the cause of the pump trip, and if corrected, within 30 minutes, attempt to restart the RCP.
.(b) trip the reactor.
(c) -shutdown the plant prior to attempting a restart of the RCP.
(d) continue operation with an upper limit of 35% reactor power.
7.15 During refueling operations the refueling canal level begins to (1.0) decrease. The operator on the manipulator crane has just started
~
withdrawing a fuel assembly from the upender when the manipulator crane radiation monitor alarms. The fuel handling SRO declares an evacuation of the reactor building and notifies the control room. During personnel accountability the crane operator is missing and is believed to be in the reactor building. It is ascertained that any_re-entry of the reactor building will give twice the 10 CFR 20 whole body exposure limit. An entry (a) cannot be made for a search because the dose rate is too high.
i (b) can be made if the Emergency Director approves of exceeding the 10 C.FR 20 limit.
(c) can be made if the Health Physics Technician approves of exceeding the 10 CFR 20 limit.
(d) can be made if the fuel handling SRO approves of exceeding the 10 CFR 20 limit.
~
26 1
l 7.16 Which of the following is true concerning a-liquid waste release? (1.0)
(a) The release cannot be made until all parts of the Liquid Waste Release Worksheet are returned to the main control room.
(b) A channel check and a source check must be performed on the radiation monitor being_used.
(c) After-a' sample is taken from the waste monitor tank, up to 100 gallons may be added without taking a new sample.
(d) A release cannot be made when the applicable radiation monitor is inoperable.
7.17 List the immediate operator actions for abncrmal nuclear power (1.0) generation when expecting a reactor trip.
7.18 Which of the following statements describe the RCP trip criteria (1.0) following an SI with normal containment conditions?
(a) Less than 23 degrees subcooling by core exit T/Cs and Pzr level less than 7%.
(b) RCS pressure less than 1380 psig and at least* one charging pump providing flow.
(c) RCS pressure less than 1600 psig and less than 23 degrees subcooling.
(d) RCS pressure less than 1600 psig and Pzr level less than 7%.
7.19 Which of the following is true of the Natural Circulation (1.0)
Emergency Operating Procedure ?
(a) Whenever a reactor coolant pump can be started, RCP 8 has first priority for restart.
(b) RCS Tc temperatures are plotted on a temperature vs Pressure graph.
(c) The preferred method of RCS depressurization is throu.,h the reactor vessel head vent.
t (d) The cooldcwn rate is limited to 50 F/hr.
a
27 7.20 Which of the following'does NOT provide adequate guidance to (1.0) confirm ~ adequate core cooling?
(a) Core exit thermocouples.
(b) RCS Th temperatures.
.(c) _RCS loop AT.
(d) Narrow range reactor vessel level indication.
7.21 List the immediate actions of the Reactor Trip or Safety . (3.0)
Injection procedure.
7.22 Which of the following is a correct response with uncontrolled (1.0) depressurization of all steam generators?
(a) Increase EFW flow to maintain SG 1evels greater than 38%.
(b) Locally unlock and close isolation valves for any failed SG code safety valves.
(c) Shut'all SG main feed and main steam isolation valves.
(d) Isolate any RCS borations in progress to prevent further RCS cooldown.
7.23 Which of the following is appropriate operator action for a steam (1.0) generator tube rupture?
(a) After verifying the SG with the tube rupture has its' main steam line isolated, break the condenser vacuum.
(b) If all RCPs are stopped, then RTO bypass manifold temperatures should be used for RCS cooldown indication.
(c) Stop the TDEFW pump when the SG Level Low alarm citars.
(d) If condensate storage tank level falls to 12 ft., shift-the feedwater pump suction to service water.
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. , - - ,r n- ~~ - r , * -
+,
28 7.24 List the immediate operator actions for a loss of all AC power (1.0) 7.25 The procedure used for recovering-from a SGTR when minimizing (1.0) of RCS inventory loss is of primary concern is (a) SGTR with loss of Reactor Coolant Subcooled Recovery (b) SGTR with loss of Reactor Coolant Saturated Recovery (c) SGTR without Pressurizer Pressure Control (d) Steam Generator Tube Rupture 7.26 The Loss of Reactor or Secondary Coolant Procedure directs the (1.0) operator to establish hot leg recirculation (a) 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> after the start of the event, and then every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> cycle between hot leg and cold leg recirculation lineups.
(b) as determined by the TSC staff, and then cycle between hot leg and cold leg recirculation. lineups every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(c) 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> after the start of the event, and then cycle between hot leg-and cold leg recirculation lineups as required by the results of RCS boron concentration sample analysis.
(d) when the RWST reaches its low level alarm setpoint, and then cycle between hot and cold leg recirculation lineups as determined by the TSC staff.
7.27 List the V. C. Summer Administrative exposure limit for whole (1.5) body per quarter, whole body per year, and extremity per quarter.
7.28 If you receive ik Rems of whole body exposure at V. C. Summer (0.5)
Nuclear Plant in a calendar quarter when you have scheduled an annual chest x-ray, you are required by 10 CFR 20 to reschedule your x-ray to another quarter. True or False?
7.29 Which of the following readings is the maximum allowable for (1.0) entering the RCA, as read on a 0-500 mr self reading dosimeter, without requiring HP to zero it?
(a) 200 mr (b) 250 mr (c) 300 mr (d) 350 mr SECTION 7.0 IS CONTINUED ON NEXT PAGE
29 7.30 When frisking upon exiting a radiological controlled area (1.0)
(a) ensure the frisker is set on the X5 scale.
(b) the frisker should be held one inch away from your body.
(c) you should stop the frisker probe momentarily over an area you notice an increase in count rates.
(d) immediately leave the area if the frisker alarms.
- WRITE "END OF SECTION 7.0" ON YOUR ANSWER SHEET.
30 8.0 Administrative Procedures, Conditions, and Limitations (32.0) 8.1 Which of the following statements is true concerning red (1.0) danger tags?
^
(a) Any qualified Danger Tagger can perform a second verifi-cation of tag placement provided he did not perform initial placement.
(b) A component cannot have a danger tag and a caution tag attached to it. If the situation occurs, the caution tag .
must be cleared.
(c) The green copy of the work request is filed in the active Danger Tag Log after'the tags are hung.
(d) Electrical personnel may act in the capacity of a red tag.
8.2 In Modes 1, 2, 3, and 4j Technical specifications require (1.0) as part of the minimum shift crew.
(a) Two Reactor Operators, one auxiliary operator and one STA.
(b) One Reactor Operator, one auxiliary operator and one STA.
(c) Two Reactor Operators, two auxiliary operators and two STAS.
(d) Two Reactor Operators, two auxiliary operators, and one STA.
8.3 When trouble-shooting, a shall be written to (1.0) restore the equipment to ensure all recorded data and inspection requirements are maintained.
(a) Maintenance Work Request (b)' Preventative Maintenance Task Sheet (c) Shop Work Order (d) Surveillance Test Task Sheet i
, - . - , , - . , -n-- ,, - - , , - --0---
31 8.4 The operations Key Control Log Books consists of: (1.0)
(a) security keys and high radiation area keys (b) equipment keys and high radiation area keys
-(c) security keys and equipment keys (d) security keys and non-security keys /non equiptent keys.
8.5 Which of the following does NOT have unrestricted access (1.0) to the control room?
(a) H. P.-Supervisor (b) STA (c) NRC Resident Inspector (d) A & B NR0s 8.6 If an operator scheduled for the next shift calls in sick, the (1.0) absent operator should be replaced by:
(a) calling in an operator on the administrative shift.
. (b) keeping the duty operator over 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and calling the operator that would have relieved the absentee operator in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> early.
(c) calling in an operator on the Training Shift.
4 (d) calling in an operator on vacation if known to be in the area.
8.7 List 4 Logs that are required to be reviewed prior to a shift (1.0) turnover by the oncoming control room sopervisor.
8.8 The process of determining an instruments accuracy by (1.0) visually comparing the indication to other independent instrument channels measuring the same parameter is defined in Tech Specs as a (a) Channel Calibration.
(b) Channel Check.
(c) Channel Functional Test.
(d) Channel Source Check.
, . - ,, - ,, -- .,-,v-.- ,, ,- - - -- - h- - - , - - - - , - - - , , . --
32 8.9 Match the RCS leakage types in Column A to the Technical (1.5)
Specification limits in Column B.
Column A Column B
- 1. Pressure Boundary a. O gpm
- 2. Identified 'b. 1 gpm
- 3. Unidentified c. 20 gpm
- 4. Controlled d. 10 gpm
- 5. Primary-to-secondary e. 33 gpm 8.10 Which of the following is true concerning the Temperature, (1.0) -
Power, and Pressure Reactor Safety Limit Curves? .
(a) As Tavg decreases, the acceptable press'urizer pressure increases for a ccnstant power level.
(b) As power level increases, the acceptable pressurizer pressure increases for a constant Tavg.
(c) The curve is based on three reactor coolant pumps running.
(d) The curve is based on three reactor coolant pumps running and three steam generators MSIV open.
8.11 Withdrawal of the shutdown rods following a reactor trip -(0.5) has no effect on the available shutdown margin. TRUE or FALSE?
8.12 The sh'utdown margin must be greater than or equal to % (1.0) delta K/K-in Modes 1 and 2 and greater than or equal to % delta K/K in Mode 3, 4 and 5.
8.13 According to the Technical Specifications, three conditions (1.0) must exist for the RWST to be operable. Which of the following is NOT one of these three conditions?
(a) A maximum water temperature.
(b) A minimum water volume.
(c) A minimum boron concentration.
(d) A minimum water temperature.
33 8.14 If Tavg drops below 551 'F, Tech Specs allow (1.0) to restore it to 551 *F or greater when in Mode 1.
(a) 15 minutes (b) 30 minutes (c) _1 hour (d) 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> 8.15 If less than the minimum required containment smoke (1.0) detectors are operable, (a) the containment air temperature must be monitored at least once per hour.
(b) a fire watch patrol inspection must be conducted in containment at least once per hour.
(c) action must be initiated to place the unit to at least HOT STANDBY within one. hour, if operating.
(d) no action is required other than to prepare and submit a Special-Report to the Commission within the.next 30 days.
8.16 If control power is lost to a pressurizer power operated (1.0) relief valve while in Mode 1, (a) no action is required by Tech Specs provided another PORV is operable and all pressurizer code safety valves are operable.
-(b) Tech Specs require the power supply to be removed from the associated block valve after verifying it to be open, if the PORV is not made operable within one hour, to continue to operate.
(c) Tech Specs require the associated block valve to be shut-and its power removed if the PORV is not made operable within one hour, to continue to operate.
(d) Tech Specs require action to be initiated within one hour to place the plant in at least HOT STANDBY within the following hour, if the PORV is not made operable.
34 ;
\
8.17 The LC0 limiting primary containment average air temperature (1.0) l to be less than 120 *F is based on (a) the dew point for the Tech Spec containment pressure band.
(b) the component cooling water providing adequate cooling to loads inside containment.
(c) the temperature assumed in the accident analysis for a LOCA or steam line break inside containment. ,
(d) temperature required for proper operation of instruments .
inside containment.
8.18 In what modes are the containment hydrogen mitigation (1.0) systems required to be operable by Tech Specs?
(a) Mode 1 j' (b) Modes 1 and 2 I~' (c) Modes 1, 2, and 3 (d) Modes 1, 2, 3, and 4 8.19 In Mode 6, Tech Spec LCOs require (1.0)
(a) audible source range indication in the containment but not the control room.
(b) normally at least one RHR loop in operation.
(c) the containment purge and exhaust isolation system in operation.
(d) direct communication maintained between the reactor building and the spent fuel storage area.
8.20 Which of the following is NOT required of a.c. electrical (1.0) power sources by Tech Specs in Modes 5 and 67 (a) Two circuits from the offsite transmission network to the on site Class IE distribution system.
(b) One diesel generator.
(c) Three 120 volt vital Busses aligned to their associated inverters.
(d) Three 480 volt A. C. Emergency Busses.
35 8.21 Name the North and South holding areas that non-essential (1.0) personnel are evacuated to when required by an emergency.
8.22 Fire Brigade Leader tasks are performed by (1.0)
-(a) Shift Foreman (b) Shift Supervisor
.(c) The "B" Nuclear Operator .
(d) Emergency Coordinator .
8.23 When handling a contaminated and injured person, it is NOT (1.0) required (a) to have a central security control escort of outside ambulance service.
(b) to have the injured person to be decontaminated prior to being transported to a hospital.
(c) for off-site ambulance personnel to be issued dosimetry.
(d) to notify the NRC of the injured person being transported to the hospital.
8.24 In the event of a General Emergency, the general population (1.0) down wind of the V. C. Summer Nuclear Plant may be recommended to evacuate up to from the plant.
(a) 2 miles (b) 5 miles t
(c) 7 miles (d) 10 miles 8.25 The least Severe emergency classification that requires (1.0) activation of the Technical Support Center is (a) an unusual event.
(b) an alert.
(c) a site area emergency.
(d) a general emergency.
36 8.26 Which of the following statements is true concerning a (1.0) temporary procedure change approval?
(a) The Discipline Supervisor does not need to review a temporary change.
(b) The Shift Supervisor must approve a temporary change prior to final approval.
(c) the sequential change number for a temporary change is '
from a separate source than that of a permanent change. -
(d) The Procedure Development Form must be placed in the
~
control room copy of the procedure when distribution of the change is made.
8.27 The Station Log Book, the R&R Log Book, the BISI, and the (1.0).
Status Board must have a formal audit performed to ensure they are all in agreement as to the safety system status at least once per s
(a) shift.
(b) day.
(c) week.
(d) month.
8.28 List four required log entries when assuming the watch.for (1.0) day shift.
8.29 Which of the following is NOT a required one hour or a (1.0) four hour reportable event?
(a) The initiation of a plant shutdown in accordance with a Tech Spec action statement.
(b) A routine radioactive waste gas release.
(c) A deviation from Tech Specs required in an emergency situation to protect the public health and safety when there is no action provisions in the license conditions or Tech Specs to adequately provide the protection.
(d) Transporting a contaminated person to the Richland Memorial Hospital.
. l.
37 8.30 Which of the following is contrary to the Code of (1.0)
Federal Regulations concerning licensed personnel?
(a) Senior operator licenses expire two years after the date of issuance.
(b) The NRC must be notified of any medical disability incurred by an applicant for a Senior Operators license or by a licensed senior operator.
(c) An unlicensed individual may manipulate the controls .
of the reactor as part of his training to qualify for .
an operator license if under the direction and in the presence of licensed operator.
i-(d) -The Shift Supervisor may assume the duties of Fuel Handling Supervisor during core alterations when it is decided that the alterations can be supervised best from the main control room.
8.31 A standing radiation work permit will NOT tell you (1.0)
(a) the dress requirements.
(b) the dosimetry requirements.
(c) your stay time.
(d) if respiratory protection is required.
8.32 Which of the following is a 10 CFR 20 exposure limit? (1.0)
(a) 1000 mrem / quarter whole body.
(b) 5 rem / year whole body.
(c) 15 rem /qtr to extremities.
(d) 1250 mrem /qtr of beta radiation exposure to the lens of the eye.
WRITE "END OF SECTION 8.0" ON YOUR ANSWER SHEET.
m.
-e T ABLE D-lb Properties of Dry Saturated SteamIcontenued.
Temperature Enthairs ! Entropi specirec solume Temp.. A h' Sat 54: Sat
'F P" - Sat 54: San hqw.d E'*r *,por ' l hovid 1
l Es p ,,p ,,
P houid sapor 4, 4 ,, 4, ! i, s, r, e e i, i, 0.00 10758 1075.8 0.0000 2.1877 2.1b77 32 00M854 0 01602 3306 1074.1 1077.1 0.006l 2.1709 2.1770 3.02 0.09995 0 01602 2947 ',
35 2444 8 05 1071.3 1079.3 0 0162 2.1435 2.1597 40 ' O.12170 0 01602 0.0262 2.1167 2.1429 2036 4 13.06 1066 4 1081.5 45 014752 0 01602 0.0361 2 0903 2.1264 1703 2 18 07 1065 6 1083.7 50 0.17811 0 01603 1059.9 1088 0 0.0555 2.0393 2.0948 0.2563 0 01604 1206.* 28 06 60 I9902 2.0641 38.04 1054.3 70 0.3631 0.5069 0.01606 0 01608 867 9 633 1 48.02 1048 6 1092.3 1096.6 0.0932f. 1.9426 0.0745 ' 2.0360 80 1042.9 1100.9 0.1115 1.8972 2.0081 0.6962 0 01610 468.0 57.99 90 1105.2 0.1295 1.8531 l.9826 0.01613 350.4 67.97 1037.2 100 0.9492 77.94 1031.6 1109.5.01417 1.ltIO6 f.9577 110 1.2748 0.01617 265 4 1.9339 87.92 1025.8 1113.7' O.1645 1.7694 120 1.6924 0 01620 203.27 1020.0 1117.9 0.1816 1.72 % 1.9112 2.2225 0.01625 157.34 97.90 130 1122.0 0.1984 1.6910 1.8894 0.01629 123.01 107.89 1014.1 140 2.8886 l.6537 1.8685 97.07 117.89 1008.2 1126.1 0.2149 150 3.718 0.01634 1130.2 0.2311 1.6174 1.8485 0.01639 77.29 127.89 1002.3 160 4.741 1.8293 62.06 137.90 996 3 1134.2 0.2472 1.5822 170 5.992 0.01645 1.8109 147.92 990.2 1138.1 0.2630 1.5400 ISO 7.510 0.01651 50.23 984.1 1142.0 0.2785 1.5147 1.7932 9.339 0.01657 40.96 157.95 190 977.9 1145 9 0.2938 1.4824 l.7762 11.526 0.01663 33.64 167.99 200 971.6 1149.7 0.3090 1.4508 1.7598 0.01670 27.82 178.05 210 14.123 1.4446 1.7566 26 80 180.07 970.3 1150 4 0.3120 212 14.6 % 0.01672 1.4201 1.7440 23.15 188.13 %5.2 1153.4 0.3239 220 17.186 001677 1.3901 1.7288 19.382 198.23 958.8 1857.0 0.3387 230 20.780 0.0i684 1.3609 1.7140 16.323 208.34 952.2 1140.5 0.3531 240 24.% 9 0.01692 1864.0 0.3675 1.3323 1.6998 0.01700 13.821 216 48 945.5 250 29.825 1.3043 l.6860 l 1.763, 228.64 938.7 1167.3 : 0 3817 260 35 429 001709 f.2769 f.6727 10 061 l 238.84 931.8 1170 6 " 0.3958 270 41.858 001717 f.2508 1.6597 8.645 249 06 924.7 1873 8 0.4096 280 49.203 0.01726 1.2238 1.6472 7.46l 259.31 917.5 1176.8 0 4234 290 37.556 0.01735 1.1980 1.6350 67.013 0.01745 6 466 269.59 910.1 1179.7 0 4369 300 0 4504 1.1727 f.6231 310 77.68 0.01755 5.626 279 92 902.6 1182.5 1.1478 1.6115 89 66 0.01765 4.914 290.28 894 9 1185.2 0.4637 320 1.1233 1.6002 330 103 06 0.01776 4.30' 300 68 887.0 1887.7 0 4769 1.0992 f.5891 118.01 0.01787 3.788 311.13 879 0 1190 1 0 4900 340
c TABLE D-lb Properties of Dr5 Saturated Steam tenntmurd.
Temperature
$re.ifes wiume k nihairy knirors g
Temr. p,,,,
- F 54: Sat bi Sai Sai Sat I* hqu4 separ hqu4 Esar s arm liquid b'*I serM f i, i, 4 4, 4, s, ., i, .
350 134 63 001799 3.34 321.63 870 7 1192.3 0.M:9 1.0754 1.5783 '
360 153 04 0 01511 :.95- 33: 18 852.2 1194 4 0.5158 1.0519 1.5677 310 173.37 0 0182.2 ': 625 34: '9 853 5 1196 3 05286 10:b7 1.55'3 3k0 195.7* 0 0lb36 2 335 353 45 1544 6 1898 1 0.5413 1.0059 1.5471 390 2 0.37 0 01150 2.06.26 364 17 8354 1899 6 0.5539 0 9832 1.5371 400 247.31 0 01864 1.6633 374 97 826.0 1201 0 0.5664 0.9608 1.527 410 276.75 0.01876 1.6700 38583 816.3 12011 0.5788 0.9386 1.5174 4:0 308.83 0 01894 1.5000 3 % 77 806 3 1203.l 0.5912 0.9166 1.5078 430 343.72 0.01910 13499 407.79 7% 0 1:03 8 0 6035 0.8947 1.4982 440 381.59 0.01926 1.2171 418 90 785.4 , 1204.3 0.6156 0 8730 1.4887 450 422.6 0.0194 10993 4301 774.5 1204 6 0 6280 0.8513 1.4793 460 466 9 0 01 % 09944 441 4 76) 2 I:04 6 0.640 0.8298 1.4700 470 514.7 00198 0.9009 4518 751 5 1204 3 0.6523 0.0083 14606 460 5661 0 0:00 0817: 464 4 739 4 I:03 7 0 6645 0 7868 1.4513 490 621.4 0 0:02 0 74:3 476.0 726 8 1:011 0 6766 0.7653 1.4419 500 6806 0 0:04 06140 4 tit 8 'l39 1:01 ? O6as- 07438 143:5 '
5:0 812.4 0 0:09 0.5594 511.9 686 4 Ilve : ; 0 7 30 0.?006 1.4136
$40 962.5 0 0215 0.4649 536 6 656 6 1193 : !07374 0.6568 1.3942 560 1I33.1 0 02:1 0.3666 562.2 624.2 1186 4 0 7621 0 61:1 I.3742 ,
$80 1325.8 0 0:28 0.3217 586.9 588 4 I I77.3 0.787 0.5659 1.353:
600 1542 9 0.0236 0 2668 610.0 548.5 1165 5 0.8131 0 5176 1.3307 620 1786.6 00:47 0.2201 646.7 503 6 II50.3 0 8398 0 4664 1.3062 640 2059.7 0 0260 0.1798 678 6 452.0 1130.5 0 8679 0.4110 1.2789 660 23654 0 0278 0 144 714.2 390 2 1104 4 0 9987 0.3485 1.2472 650 2708 1 0 0305 0 til! 757.3 309.9 1067.2 0.9351 0:719 1.2071 700 3093 7 0 0369 0 0761 823.3 172.1 995 4 0 9905 0.1484 1.1389 705 4 3:06. 0 0503 0 0503 902 7 0 902.7 1 0560 0 1.0580 4
i 4
9
1 1 A RI T D.I.i' Properties of Dr) SJiuf.Ild! hicJm
- Prenure spud.. . som.
3, fTemp E ninair> E ntrop>
pres.
Es.' 54: Sai 54i Sai 54: 54:
T liqu d vapor ligued EF sapor I"P leque ..por ,
r u , s, ha h, h, n, s, s, 1.0 101.74 0.0 14 333 6 69 70 10 %.3 1106.0 0.1326 1.84 % 1.9782 2.0 126 08 0 01023 173 73 93 99 1022.2 I i 16.2 0I149 1.745I l.9200 30 1414k 0 01630 118.71 100 37 1013.2 l122.6 02008 1.6855 1.8863 40 152.97 0.016 % 90 63 12ti h6 1006 4 l l27.3 0.2198 1.6427 1.8625 5.0 162.24 0 01640 73.52 130 13 1001.0 l 131.1 0.2347 2.6094 I.8441 60 170 06 0 01644 61.98 1.17.% 996.2 1134 2 0.2472 1.5820 1.8292 7.0 176 85 0 01649 53 64 144 76 992.1 l 136.9 0.2581 1.5586 1.8167 80 182.86 0 01653 47.34 150.79 988.5 1139.3 0.2674 1.5383 1.8057 90 188.28 0 016 % 42.40 156 22 985.2 1141.4 0 2759 1.5203 1.7962 10 193.21 0.01659 38 42 161.17 982.1 1143.3 0.2835 1.5041 1.7876 14.696 212.00 0.01672 26 80 180.07 970.3 1850.4 0.3120 1.4446 1.7566 15 213.03 0.01672 26.29 ist 11 %9.7 1150.8 0.3135 1.4415 1.7549 20 227.96 0.01683 20 089 196 16 960.1 l156.3 0.33 % 1.3962 l.7319 25 240.07 0 01692 16.203 208 42 9521 1160.6 0.3533 1.3606 1.7139 30 250.33 0 01701 13 746 218.82 945 3 1164.1 0.3680 1.3313 1.6993 35 259.28 0.01708 II.896 227.91 919.2 1167.1 0 3807 1.3063 1.6810 40 267.25 0.01715 10.498 236.03 933.7 1169.7 0.3919 1.2844 1.6763 45 274.44 0.01721 9 40i 243 36 928.6 1872.0 0.4019 1.2650 1.6669 50 281.01 0.01727 8.515 22 09 924 0 1874.1 0.4110 1.2474 1.6585 55 287.07 0.01732 7.787 256.J0 919.6 1175.9 0 4193 1.2316 1.650#
60 292 71 0 01738 7.175 262.09 915.5 1177.6 0.4270 1.2168 1.6438 65 297.97 0 01743 6 655 267.50 911.6 1179.1 0.4342 1.2032 1.6374 70 302 92 0 01748 6.206 272.61 907.9 1180.6 0 4409 1.1906 1.6315 75 307.60 00153 5 816 277.43 9045 1881.9 0 4472 1.1787 1.6259 80 312 03 0.01757 $472 282.02 901.1 1683 1 0 4531 1.1676 1.6207 85 316.25 0 01761 5.168 286.39 897.8 Ilb4 2 0 4587 1.1571 1.6158 90 320.27 0 01766 48% 290.56 894 7 1185 3 0 464I l.1471 1.6112 95 324 12 0.01770 4 652 294.56 891.7 1186.2 0.4692 1.1376 1.6066 100 327 81 0 01774 4 432 298 40 888 8 l i b7.2 0 4*40 1.1286 1.6026 110 334 77 0.01782 4 049 305 66 883.2 1886 9 0 4532 1.1117 1.5945
r-O TABLF D la Properties of Dr> S4iurated Sicam tronnnueds Pressurc i nircry
$pearic solum Entheir) g Temp.
set $4s Lt F"^ F $4: 54 Ses I'*P sapor hque I'dP ' s aro, hauid sapor hwd 4, +. e,
., i, 4, 4 ,, ..
a r
1190 4 0 4916 10962 1.587b 120 341.25 0 01789 37k 312.44 877.9 I191.7 0 4995 1.0817 l.581:
3 455 318 81 872.9 '
I30 347 32 0 01790 868 2 Il93 0 0 5069 1068: 1.5751 ,
3.220 324 82 140 353 02 0 0180: 863 6 1l94.1 0.513k 1.05 % 1.5694 IN 35k 4: 0.01809 3.015 330SI 1195.1 0.5204 1.0436 1.5640 2.834 335.9.1 859.2 160 363.5) 0 01815 1.5590 0 0162: 2 675 341.09 854 9 It%0 0.526e 1.03:4 170 166 41 k50 k l196 9 0 53:5 1.0217 1.554:
0 01627 < vt32 346 01 180 373 06 B4o n 18976 0.5381 1.0llo 1.5497 0.01833 2 404 350 79 190 377.51 1198 4 0 5435 1.00l* 1.5453 0.01839 2.2p 355.36 843 0 200 381.79 1201.1 0.5675 0.9588 1.5263 400.95 0 01865 1,8438 376.00 825 l 250 1.5104 809 0 12018 0.5879 0 92:5 417.33 0 01890 15433 393 84 300 1203.9 0.60 % 0 8910 1 4966 0 01913 1.360 409 69 794 2 350 431.72 0 6:14 0.8630 1.4844 0.0193 11613 424 0 7805 I:04 5 400 444.59 0 63 % 0.8378 l.4734 1.03:0 437.2 767.4 1:04 6 450 456.28 0 0195 0.8147 1.4634 09 78 449 4 755 0 1204 4 0.6487 500 467.01 0.0197 460.8 743 l 1:039 06608 0 79.4 1 1.4542 550 476 94 0 0199 0.8424 1.4454 471.6 731.6 1203 2 0 6720 0.7734 600 486.21 0 0:01 0.7698 1.4374 481.8 7:0.5 1202.3 0.6826 0 7548 650 494.90 0 0:03 0.7083 1.42 %
491.5 709.7 1:01.2 0 69:5 0.7371 700 503 10 0.0 05 0 6554 14:23 699.2 1200.0 0 70)9 0.7204 750 $10 86 0 0:07 0609: 500 8 1898 6 0.7108 0.7045 1.4153 0 0209 0 % 87 509.7 688 9 800 $18.23 0.7194 0.6891 1.4085
$18 3 678.8 1197.1 850 525 26 00:10 0 5327 0.7275 0 6*44 1.4020 0 5006 526 6 668.8 1195.4 900 531.98 0 0:12 0 660: 1.3957
$34 6 659.1 1193.7 0.7355 950 538 43 0 0:14 0 4717 1.3897 649 4 1191.8 07430 0.6467 4461 0.0:16 0 44 % 542.4 1000 II87.7 0 7575 06205 1.3780 00:20 0 4001 557 4 6304 1100 5 % 31 0 77II 0.5936 1.3667 0 01:3 0 3619 $71.7 6Il.7 1183 4 1200 567 22 1178 6 0 764n 0 5719 I3559 0 02 ~ 0.3293 $85 4 593 :
1300 577 46 0 7963 0 5491 1.3454 0.301: 598.7 574.? 1873 4 1400 587.10 0 0:31 0 5269 1.335l 0 2765 611 6 !!6.3 1167.9 0 804:
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- 4. 1850 4 2.1720 2.2233 2.2702 2.3137 2.3542 2 .W21 2 4283 2.4625 2.4952 2.5%6 (101.74, s. 2.05l2 2.1153 102.26 114 22 126.16 138 10 150 03 161 95 173 s7 185.79 197.71 221.6
,. . . . 78.16 90.25 1241.2 12880 1335 4 13s3 6 1432.7 14n2 h 1533 4 1585.1 1637.7 1745.7 5 4. . . 1848.8 1895.0 1.9942 2.04 % 2.0927 2.1361 2.1767 2 214n 2.2509 2.2851 2.3178 2.3792 (162.24, s.. . l.8718 1.9310
$104 57.05 6303 69 01 74 9N no 95 n6 92 92 8s 98.84 110.77
- v. 38.85 45.00 1240.6 1287 5 1335.1 1383 4 1432 5 14n2 4 1513 2 15n5 0 1637 6 1745.6 10 A.. 1146 6 l193.9 l.7927 l.8595 1.9172 1. % 89 2.0160 2.05% 2.1002 2 Iini 2 1744 2.2068 2.2413 2.3028 (193.21) s, 34 6s 38.78 42.56 46 94 51.00 55 07 59 83 63.19 67.25 75.37
- v. . 30.53 1239 9 1287.1 1334s 1383 2 1432.3 1482.3 15331 15n4 s 1637.5 1745.5 14 696 A .. 1892.8 1.9734 2.0170 2.0576 2.0958 2.1319 2.1662 2.1989 2.2603 (212.00) s. I8160 1.8743 1.9261 25 43 28 46 3147 34 47 37.46 40.45 43 44 46 42 49 41 55.37
,, 22.M 1239 2 1286.6 1334 4 1382.9 1432.1 1482.1 1533 0 15s4 7 1637 4 1745 4 20 A. 1891.6 2.0978 2.1321 2.164s 2.2263 ,
l.7808 1.83 % 1.8918 1.9392 1.9529 2.0235 2.0618 (227.96) s, II.040 12.628 14.168 15.68s 17.19s 18.702 20.20 21.70 23 20 24 69 27.68
- 1236.5 1284.8 1333.1 13bl .9 1431.3 1481.4 1512 4 1554 3 1637 0 1745.1 40 4. 1136.8 1.6994 1.7608 1.8140 1.8619 1.9058 1.9467 1.9850 2.0212 2.0535 2.0s43 2.1498 (267.25) S.. . . .
8.357 9 403 10 427 11.441 12.449 13.452 14 454 15 453 16 451 18.446
,. . . 7.259 1233 6 1283.0 1331.8 1380.9 1430.5 14so s 1531.9 l$nt a 1636 6 1744 8 60 A. . . 1881.6 1.9762 20luh 2.0434 2.1049 l.6492 1.7135 1.7678 1.8162 1.8605 1 %IS 1.9400 (292.711 s . .
6.220 7.020 7.797 8.%2 9.322 10 077 10.830 ll.5n2 12.332 13.830 v.. . .. . . .
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4 937 5.589 6.218 6 835 7.446 8.052 8.656 9.259 9 860 11.080 v... , .
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1.8829 1.9193 1.9538 1.9867 2.04s4 l.6518 1.7085 1.7541 l.8029 l.8443 (327.81) s .. .... ..
4.081 4.6M 5165 5.683 5.195 6.702 7.207 7.710 8.212 9.214 1377.8 1428.1 1478.8 1530.2 1582.4 1635.3 1743.9 130 h. . . . . . . . .. . . 1224 4 1277.2 1327.7 2.0281 l.8237 2.0625. l.0980 1.9335 1.9664 lut.25) s. . . .
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- v. 2 649 3N4 3 411 3.764 4.110 4 452 4.792 5.l29 5.466 6.136 180 4. . . . 1214 0 1271 0 1323 5 1374.7 1425 6 1476 8 1528 6 15s1.0 1634.1 1742.9 (373 06) s. 1.5745 1.6373 16s94 17355 1.7776 I8167 l8534 l.8882 1.9212 I9831
- 2.36l 2.726 3 060 3 380 3 693 4 002 4 309 4 613 4 917 5 521 200 4. 1210.3 1268 9 1322.1 1373 6 14248 1476.2 1528 0 1580.5 1633.7 1742.6 (381.79) s. I.5594 i6240 16767 1.7232 1.7655 1.8048 I8415 1.8763 1.9094 1.9713
- v. 2.125 2 465 2.772 3 066 3.352 3 634 3 913 4 691 4 467 5 017 220 A. 1206 5 1266 7 1320 7 1372 6 1424 0 1475.5 1527.5 1580 0 1633.3 1742.3 1389 80) s 1.5453 16117 16e52 1.7120 1 7545 I 7939 1.8308 ' 3656 f.89s7 89607
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EQUATION SHEET Q = mah RR = Iffth g y+g=
KE +h KE *h S 2 2 * *12 SCR =
Q = UAAT 1-Keff Where:
3 1) KE is Kenetic energy x = xw P CR 2) w is work done 1 3) o is the heat Q = mcpAT M = transferred Cr o
- 4) h is the enthalpy r
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t 38 LJ-5.0 Theory of Nuclear Power Plant Operation, Fluids and Thermodynamics o- ANSWERS (32.0) 5.1 (b) (1.0)
Ref: V. C. Summer Rx Theory Review Text, p.1-2.17 5.2 (b) (1.0)
Ref: V. C. Summer Rx Theory Review Text p., I-4.19 - I.4.21 5.3 (c) (1.0)
P=Po 10 SUR(t) ,
~
in (P/Po) = SUR(t) in (10)
In (5000/1000) = SUR (30) in (10) 1.6094 = SUR(30) 2.3026 1.6094 60 see SUR =
min
= 1.3979 2.3026 (30 sec)
Ref: V. C. Summer Rx Theory Review Text, p. I-3.15 5.4 (d) (1.0)
Ref: Nuclear Energy Training, Module 3, NUS Corp.
5.5 (a) (1.0)
K -K i , 1.004 - 0.92 , 0.084
.091 e
K K i (1.004)(0.02) 0.924 Ref: Nuclear Energy Training, Module 3, NUS Corp., p.6.1 5.6 (b) (1.0)
Ref: Nuclear Energy Training, Module 3, NUS Corp., Sec. 10 5.7 (b) . (1.0)
Ref: V. C. Summer Xenon Oscillation Handout, p. 4 5.8 (d) (1.0) 1 Nt = So 1-Keff
= 100 1 1 .95 = 2 x 103 cps Ref: V. C. Summer Rx Theory Review Text, p. 1-4.13
_ _ , . _ _ _ _ _ _ _ _ _ , - _..._,_r.. _. . . , - . . -
_y. . .m-.y ., . _ - . _ . - . . - - . _ , , , , - , , , _ . , , _ . _ -_
39 5.9 (b) (1.0)
Ref: V. C. Summer Reactor Theory, I-5.78 5.10 (a) (1.0)
Ref: Nuclear Energy Training, Module 3, NUS Corp.
5.11 (b) (1.0)
Ref: V. C. Summer Reactor Theory, I-5.42 5.12 (a) (1.0) '
Ref: V. C. Summer Reactor Theory, p. 1-2.30 .
5.13 (c) (1.0)
Ref: V. C. Summer Reactor Theory, p. 1-3.10 5.14 (b) (1.0)
Ref: VCS T/S, Para. 3.1.3.6, 4.2.1.4, Station Curve Book.
5.15(c) (1.0)
Ref: Nuclear Energy Training, Module 3, NUS Corp. pp.8.3-1, 9.2-5.
5.16 (a) (1.0)
Ref: VCS Reactor Theory, p. I-5.31 5.17(a) (1.0)
Ref: Nuclear Reactor Analysis, Duderstadt and Hamilton, 1976, p.13 5.18 (d) (1.0)
Ref: 10 CFR 50.46(b)(2).
5.19 (a) (1.0)
Ref: TMI, Report to Commissioner and Public Vol. II, Part 2, p. 527 .
5.20 (d) (1.0)
Ref: VCS Reactor Theory, p. 1-5.27 5.21(c) (1.0) 2000
= 1.11 1800 power = (1.11) '40 = 54.76 HEA0 = (1.11) #20 = 24.64 Ref: Heat Xfer, Thermo & Fluid Flow Fundamentals, General Physics Corp., pp.322-324 5.22 (d) (1.0)
Ref: Nuclear Energy Training, Module 4, p. 6.2-4
c.
40
.9 5.23 (a) (1.0)
Ref: Steam Tables 5.24 (b) (1.0)
Ref: VCS T/S Bases 3/4.4.9 5.25 (b) .
(1.0)
Ref: Heat Transfer, Thermo & Fluid Flow Fundamentals, 5.26 (d) (1.0)
Ref: Nuclear Energy Training, Module 4, ,
NUS Corp., 3.3-2. .
5.27 (d) (1.0)
Ref: VCS, T/S Bases 3/4.4.9 5.28 (b) (1.0)
Ref: VCS T/S, Figure 3.2-4.
5.29 (d) (1.0)
Q = m Cp (Ts - Tg) - RCP Heat + B/D Heat Rate Ref: Nuclear Energy Training, Module 4, NUS Corp., p.2.2-4.
VCS STP-102.002 5.30(a) (1.0)
Throttling process - constant enthalpy h = 1191.8 for PZR @ 1000 psig. BTU /lbm hsat for 5 psig 1131.1 superheated Ref: Steam tables.
5.31 (d) (1.0)
Ref: Heat Transfer Thermo & Fluid Flow, General Physics, Chapter 3, p. 105 5.32 (b) Second law of Thermodynamics, Heat Transfer Thermo, (1.0)
Ref: & Fluid Flow Fundamentals, General Physics
T 41 o-6.0. Plant Systems Design, Control, and Instrumentation - Answers (32.0) 6.1 (b) (1.0)
Ref: VCS Main & Reheat Steam, p. 17. W /v e5W F 2, 6.2 (c) (1.0)
Ref: VCS SGWLC , Figure IC.2.5.
6.3 '(c) (1.0)
Ref: VCS, Containment Spray, p. 19 6.4 (b) ,
(1.0)
6.5 (d) (1.0)
6.6 (c) (1.0)
Ref: VCS, PZR press & level control, Table IC-3 4 6.7 (c) (1.0)
Ref: VCS, PZR press & level control, p. 15.
6.8 M i t ef: VCS, RPS, p. 26.
(1.0) 6.9 (b) (1.0)
Ref: VCS, Incore Instrumentation, p. 11.
6.10 (a) (1.0)
Ref: VCS, Main & Reheat Steam System, p. 24.
6.11 (d) (1.0)
Ref: VCS, Cire. Water System, p. 4.
6.12 (c) (1.0)
Ref: VCS, Nuclear Instrumentation, p. 18 6.13 (c) (1.0)
Ref: VCS, Rod Pgsition Indication, p. 15.
6.14 (d) (1.0)
Ref: VCS, Temperature Indication System 6.15 (d) (1.0)
Ref: Nuclear Radiation Detection, 2nd Ed.
W. J. Price, pp. 42 - 44.
6.16 (a) (1.0)
Ref: VCS, Turbine Control and Protection, p. 22, Fig. TB5.14.
42 6.17 (c) (1.0)
Ref: VCS, Rad Monitoring System, p. 9.
6.18 (b). _ (1.0)
Ref: VCS, Rad Monitoring System, p. 11.
6.19-(a) (1.0)-
Ref: '
6.21 (c) (1.0)
Ref: VCS, RHRS, p. 13.
6.22 (d) (1.0)
,Ref: VCS, ECCS, p. 31 6.23 (c) (1.0)
Ref: VCS Condensate, p. 9.
6.24 (b) (1.0)
Ref: VCS, EFWS, p. 43 6.25 (b) (1.0)
Ref: VCS, Rod Control System, p. 37.
6.26 (c) (1.0)
Ref: VCS, Safeguard Power System, p. 22.
6.27 (a) DdZed Ref: VCS, Safeguard Power System, p. 76.
6.28 (d) (1.0)
Ref: VCS, Safeguard Power System, p. 6
- VCS, Waste Gas System, Fig. AB12.1 6.30 (d) .
(1,0)
Ref: VCS, 50P-108 e: VCS, Spent Fuel Pit Cooling System, p. 6.
6.32 (d) (1.0)
Ref: VCS, Fuel Handling pp.34, 37, & 38.
43 s
7.0 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL - (32.0)
ANSWERS 7.1 (1.0)
R : Vcs, s0P-101, p. 1-7.2 (1.0)
(g);yes, cop,AppA,p.2.
73 (1.0) kjf: ves, GOP, Appendix 0 7.4 (1.0)
RkF- VCs, GOP-3, Appendix 0 75 (1.0) kIf: vcs, GOP-3, p. 9-7.6 (1.0)
77 (1.0) klf: vcs, GOP-1, p. 15.
7.8 (1.0) klf: ves, s0P-210, p. 20.
7.9 (1.0)
R -
FNP-1-A0P-18.0, PP 1-3 7 10 (1.0) gIf: yes, E0P-10.0, pp 13-14 (1.0) 711klf: ves, s0P-306. P 15-7.12 (1.0)
R -
VCs, STP-108.001 7 13 * (1.0) g{f: vcs, E09-8.0, p. 1.
(1.0) 714k{f: ycs, s09-101, p. 62 7 16 (1.0)
RIf: vcs, EPP-020 (1.0) 7.16kIf: vcs, 50P-108
n 3 44 l A
7.17 1. Manually trip reactor 0 .25 each (1.0)
- 2. Ensure turbine trips
- 3. Ensure EFW pumps running, if required
- 4. Iniate rapid boration of the RCS.
7.18 (b) (1.0)
REF: VCS, E0P-1.0, p. 10 7.19 (d) (1.0)
7.20 (c) (1.0)
7.21 Verify 9 .2 each (3.0)
- 1. Reactor tripped
- 2. Turbine tripped
- 3. At least one train of ESF buses energized
- 4. Check if SI actuated
- 5. All MFW pumps tripped.
- 6. Ensure SI Pumps running.
- 7. EFW Pumps Running
- 8. At least two CCW pump running
- 9. At least two SW pump running in fast
- 10. 2 SW Booster Pumps started
- 11. Check feedwater isolation
- 12. Containment Phase A
- 13. Ensure'RBCUs running in slow
- 14. main steam line isolation status
- 15. Check' Containment Pressure REF: VCS, E0P 1.0, pp. 4-5 7.22 (c) (1.0)
7.23 (c) (1.0)
7.24 1. Manually trip reactor 0 .25 each (1.0)
- 2. Verify turbine trip.
- 3. Isolate RCS.
- 4. Verify total EFW flow greater than390 gpm.
7.25 (b) (1.0)
REF: VCS, E0P 4.3.
r 1 45 l 7.26 (a) (1,0)
REF: VCS, E0P-2.0, pp. 19-20 7.27 whole body /qtr - 1 Rem (1.5) whole body / year - 4 Rem Extremity /qtr -12 Rem REF: -VCS, HP Manual, p. 5-13 7.28 False (1.5)
REF: 10 CFR.20.107 7.29 (b) (1.0)
REF: VCS, Health Physics, p. 5-17 7.30 (c) (1.0)
4
n 46 o
8.0 Administrative Procedures, Conditions and Limitations - (32.0)
Answers 8.1 (d) (1.0)
8.2 (d) (1.0)
Ref: VCS, T/S, p. 6-4 8.3 (a) (1.0)
8.4 (c) (1.0)
8.5 (a) (1.0)
Ref: VCS, SAP-200, p. 26 (1.0)
- 2. R &.R Log
- 3. Tagout Log
- 4. Station order & Special Instruction Log Ref: VCS Turnover Sheet 8.8 (b) (1.0)
Ref: VCS, T/S, 1.6 8.9 1. a (0.3)
- 2. d (0.3) ;
- 3. b (0.3)
- 4. e (0.3)
S. b (0.3)
Ref: VCS, T/S/ p. 3/4 4-19 8.10 (c) (1.0)
Ref: VCS, T/S, Figure 2.1-1.
8.11 True (0.5)
Ref: VCS, T/S 1.28 8.12 1.77 (0.5) 2.0 (0,5)
Ref: VCS, T/S 3.1.1.1, 3.1.1.2 8.13 (a) (1.0)
Ref: VCS, T/S 3.1.2.5 8.14(a) (1.0)
Ref: VCS, T/S 3.1.1.4
r 47 a
8.15 (a) (1.0)
-Ref: VCS, T/S 3.3.3.7 8.16 (c) (1.0)
Ref: VCS, T/S 3.4.4 (1.0) ef: VCS, T/S Bases 3/4.6.1.5 8.18 (b) (1.0)
Ref: VCS, T/s 3.6.5.1, -2. .
8.19 ( 1.0' ) -
R : VCS, T/S 3.9.2, .5, -7.1, .8
- (1.0)
Re : VCS, T/s 3.8.1.2, 3.8.3.2 8.21 North - Monticello Reservior Recreational Facility (IJ)
South - Nuclear Training Center .
Ref: VCS, EPP-012, p. 6 8.22 (c) (1.0)
Ref: VCS, SAP-200, p. 8 8.23 (b) (1.0)
Ref: VCS, EPP-009.
+
(1.0)
R : VCS, EPP-001, Attach II
.=
8.25 (b) (1.0)
Ref: VCS, EPP-016, p. 3 8.26 (b) (1.0)
Ref: VCS, SAP-139, ATT. IV (1.0)
- 2. Rx coolant pressure
- 3. RX coolant temperature
- 4. Operators on duty
Ref: 10 CFR 50.72
'48 8.30 (d) (1.0)
Ref: 10 CFR 50.54 (m)(2)(iv), 55.9, .32, -41 8.31 (c) (1.0)
Ref: VCS, HP Manual, p. 13-8 & -9.
'8.32 (d) (1.0)
Ref: 10 CFR 20.101 5 4
h ENCLOSURE 4
SUBJECT:
NRC RO Licensing Examination f DATE: February 25, 1985 V. C. Summer Nuclear Education and Training Exam Review j I. Question 2.21:
Answer "C" is incorrect for this question as per answer key given. Answer "D" is correct for this question as stated.
References:
- 1. GIA Drawing D-802-004 II. Question 3.07 and 3.08:
Both these questions deal with COPS which is an obsolete system and does not exist. COPS has been replaced by RHR suction relief overpressure protection. These questions should be deleted.
References:
- 1. Technical Specification 3/4.4 34; Amendment No. 26 dated 9/24/84
- 2. MRF-20249
- 3. Drawings: a) E-210-208 b) E-210-207 c) S-212-082 III. Question 3 14 This question has parts A-E, but the answer key has parts A-F.
Due to the fact that the point value of the answers for A-F were based on .5 points each, we request that Sections A-E be ,
reassigned point valuesLof .6 points each. Thi's would keep the total point value for the question the same. This question only has 5 parts and not 6 parts; therefore, needs to have point values reassigned to .6 each.
t
~
i ,
e IV. Question 3 20-We request.that you delete the question based on the following:
Depending on the interpretation of the question, there are 3 correct choices for this. question. Per Attachment IV-10.18, steam: generator mass goes down from 0-30% power'then ramps up asEsteam generator level camps up from 30%-100% power.
- 1. ' Answer 1: .If steam generator level camps up from 30%-100%
. power, then-steam generator mass increases from 30% to 100% power.
- 2. Answer 2: Taking the.other view that you are ramping down from 100% to 30%-power, then steam generator mass decreases from 100% to 30%.
- 3. Answer 3: If you look at just 0-30% power, then steam generator level remains at 38% with mass ramping down to a-low point at 30% power.
Conclusion:
The question is very general in what it wants.
The question asks how steam _ generator mass changes over the range of steam generator level ramp. It depends on whether you ramp up or down. Also, the only part that steam generator level is ramped is from 30%-100% power.. From 0-30% power, steam generator level is constant'at 38% with steam generator mass ramping.
References:
- l. Attachment IV-10.18
- 2. IC-2 Steam Generator Water Level Control Figure 102.2 (Steam Generator Program Level)
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Subject:
NRC SR0 Licensing Examination Date: February 25, 1985 V. C. Sunrner Nuclear Education and Training Exam Review I. Question 6.8:
As of April 19, 1984, the Automatic Reactor Trip Breaker actuation signals will deenergize the U.V. coil and energize the shunt coil associated with each reactor trip breaker. Answer "C" is incorrect and Answer "A" is a correct answer as the Automatic trip signal will also energize the shunt coil.
A.
References:
- 1. MRF 20208
- 2. GAI Drawing: IMS-43-003 IMS-43-002 IMS-43-189 II. Question 8.18:
We request that this question be deleted on the following bases.
This question is in conflict with SAP-205 which doesn't require mode memorization as_this question would have us do. This Tech.-
Spec. item is a 30 day action requirement.
A.
Reference:
SAP-205 6.0 PROCEDURE 6.1 Removal from Service 6.1.1 The Shift Supervisor shall determine the following upon discovery of an IN0PERABLE component or system or upon receipt of a request to remove a component or system from service:
A. The function of the affected component or system.
The system (s) affected. How AFFECTED may be a short statement or reference to a Tag Out Sheet.
B. The Technical Specification Operability requirements for the present plant mode. ,
C. Current status of the plant and the affected system or related systems requiring its operability.
D. Time limit for restoring the component or system to service per th.e Technical Specification Action Statement.
r-This procedure doGsn't r@ quire the Shift Supervisor to
. memorize A-D above, but only to determine A-D above based on his training and knowledge of the plant.- By Station Policy Tech Spec -items in duration of < 1 Hr. should be memorized, due to this relatively short ddration of time. Question 8.18 '
would have us memorize 30 day Tech Spec items.
III. Question We request that this question be deleted on the following bases. This question has no correct answer per the selections A-D. The answer given "A" (RHR Pump) is incorrect for this question per the following:
- 1. Drawing 04-4461-0-802-036
E0P-6.1; Page 6 of 16 Step 11 (B0)
Conclusion:
You will find that D-802-036 shows an RHR pump start for both SI and/or blackout. Both E0Ps also support the fact that RHR punps start for both SI and blackout.
I V .' Question
. We request to delete this question based on the following.
This question has no correct answer per the selections A-D.
The answer given "A" (S/G "A" Press. 650 psig with S/Gs "B" and "C" at 700 psig) was not correct. Low pressure SI on 675 psig low main steam pressure is for 1/1 detectors on 2/3 S/Gs not 2/3 detectors on 1/3 S/Gs.
REFERENCES:
Drawing 1080837
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