ML20246P676
| ML20246P676 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 03/22/1989 |
| From: | Carrol C, Jordan M, Miller R, Nejfelt G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20246P645 | List: |
| References | |
| 50-254-OL-89-01, 50-254-OL-89-1, 50-264-OL-89-01, 50-264-OL-89-1, NUDOCS 8903280289 | |
| Download: ML20246P676 (233) | |
Text
{{#Wiki_filter:__ - _ _ _ _ _ _ _ _ _ _ _ _ __ U. S. NUCLEAR REGULATORY COMMISSION' REGION III Report No. 50-254/0L-89-01 Licenses No. DPR-29; DPR-30 l Docket Nos. 50-254; 50-265 Licensee: Commonwealth Edison Company Post Office Box 767 Chicago,'IL 60690-Facility Name: Quad Cities Nuclear Power Station Examination Administered At: Quad Cities Nuclear Power Station Examination Conducted: February 6-10, 1987 Examiners:j G. M. Nejfelt ud S 7 7 /89 Datt
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R. Miller 4* # J/7 * [89 Dat'e 1 pt C. Carro1
/
^
S/7'/89' 0at'e Approved By: M. Jordan, Chief Operator Licensing iJ 3/12/9V Dat6 ) Section No. 1 Examination Summary Examination administered on February 6-10, 1989 Report No. 50-254/0L-89-01)) Written Examinations were administered to three Senior Reactor Operator- j (SRO) candidates and to five Reactor Operator (RO) candidates. Operating ) Examinations were administered to four SR0 candidates and to six R0 candidates. ) Results: I All candidates passed both the written and the Operating Examination. l 1 0903280289 890322 PDR ADOCK0500g54 l L-______-__. _ _ - E
DETAILS
- 1. Examiners G. M. Nejfelt, NRC, Chief Examiner R. Miller, Sonalyst C. Carrol, Sonalyst M. Bielby, NRC (under instruction)
M. Riches, Pacific Northwest Laboratories (PNL) (underinstruction) D. Draper, PNL-(under instruction)
- 2. Exit Meeting At the conclusion of,the site visit, the examiners met with. facility representatives. The following personnel attended this exit meeting: !
Facility Representatives: R. L. Bax, Station Manager, Quad Cities R. Robey, Services Superintendent M. Rodts, Operations Training Group Leader B. Svaleson, General Instructor T. Schares, General Instructor D. Kunzmann, Quality Assurance Inspector NRC Representatives: ,
]
R. Higgins, Senior Resident Inspector G. M. Nejfelt, Operating Licensing Examiner The following items were discussed during the exit meeting:
- a. No area access difficulties were encountered, while administering the Operating Examinations, with either Security or Health Physics.
Also, the Operating Staff and Technical ~ Support Staff were very cooperative to allow the examinations to be expedited and.. ! uninterrupted. 1
- b. Training areas that'were considered noticeably improved since the last replacement examination were: j 1
i (1) Candidates showed a good knowledge level'of the flow diagrams j used to-implement the Emergency Procedural Guidances (EPGs). (2) Candidates were proficient in locating and applying material '; to answer questions during the Operating Examination. These 1 included the usage of drawings, procedures, and. reference material. i l 1 l l \ I ?.
- c. Training areas that were suggested :for future considerations were:
(1) Provide more. hands on' usage of the process computer to,the R0 candidates, since several candidates demonstrated a lack ~ of I. familiarity to operate-the-equipment; (2) Emphasize the need of an operator to verify "Immediate Operator l Actions" taken; (3) ' Teach R0 candidates an overview of refueling activities to better appreciate, activities affecting system ~ parameters. ~ This-will allow them to better understand what is occurring on the refuel floor and.to monitor affected system parameters'.from.a higher _ level of understanding.
- d. The examiner identified that in peripheral locations some material deterioration of the plant was noted. Within the first 15 minutes of a walkthrough, the Examiner identified .several deficiencies:..a clogged drain,-leaking valve, tool left after work was accomplished, and metal chain securing high pressure bottles on a stairwell tread making a safety hazard for personnel. Training is suggested for all. personnel performing walkthrough of the' plant to eliminate such conditions,
- e. A summary of isolated procedural discrepancies and concerns that' were noted are given in Attachment 1.
- 3. Examination Review d No formal written comments for either the R0 or SR0 Written Examination f were provided by the facility at the time of the exit interview on 7 February 1,0, 1989. This can be attributed to the facility . ..
pre-examination review of the written examinations in the. Regional Offices on January 30 and 31,1989; and a post examinat_ ion review on February 9, 1989. ( Attachment 1 ) [ l b a 3 'l
ATTACHMENT I Summary of isolated procedural discrepancies or concerns that were noted, while administering the Quad Cities Operating Examinations on February 6-9, 1989.
- 1. The required immediate actions for a high LPRM are different for the alarm procedure (QOA 900-5-D) than for the abnormal procedure (QOA 700-4). The immediate actions of QOA 900-5-D include some'of the subsequent actions for 00A 700-4.
- 2. The " File History Column" on the form for tracking maintenance work is no longer required by with QAP 1500-2, Revision 32. Also, the form.to record needed information is not identified by a unique form number.
- 3. All candidates stated that, prior to increasing turbine load in a controlled fashion, the expected reactor power level would be evaluated for 10 minute intervals to ensure that a constant load rate was maintained. .This should be included in QGP 1-3, " Unit Hot Standby to Power Operation," since this is how business is expected to be conducted in the plant.
- 4. The facility should consider including frame numbers for electrical breakers on motor control centers (MCCs) in procedures utilizing the '
breaker. While testing local emergency op'erations, one candidate had to search for as long as two minutes to find a particular electrical breaker on a MCC.
- 5. The "HPCI Monthly and Quality Test" procedure (QOS 2300-1) requires the-taking of data for certain parameters but does not specify an acceptable range for system operability. This data includes pump suction pressure, turbine outlet pressure, and turbine speed. If this data is used to determine pump operability, then acceptable ranges or tolerances should be specified, as is specified for the pump flow and discharge pressure.
- 6. QGA 500-6, Revision 2, "RPV (Reactor Pressure Vessel) Flooding, "
Step I-4, provides no guidance on what to do if RPV pressure cannot be increased to + 77 psig with RPV injection.
- 7. QGA 200-3, Revision 2, " Primary Containment Pressure Centrol," Step E-3, prohibits drywell spraying if drywell temperature is greater than 340 degrees F, while QGA 200-2, Revision 2, "Drywell Pressure Control,"
Step G-3, allows drywell spraying only before 280 degrees F. Therefore, it is possible to be in a temperature band between 280 and 340 degree F with no specific action directed.
- 8. Step B-3 of QGA 100-3, Rev. 2 "RPV Power Control" and Step C.2.b of QGP 2-3, " Reactor Scram" require running back the Reactor Recirculation
Attachment 1 2 Pumps to reduce reactor power. These procedures should be looked at in light of the recent power oscillation events at LaSalle County Station.
- 9. Supporting procedures are not provided with procedures for local emergency operations, such as for local operations of the Safe Shutdown Panel with a loss of D. C. Control Power.
1
U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION REGION 3 [j, , FACILITY: Quad-Cities 1 & 2 REACTOR TYPE: BWR-GE3 f;. g ; , DATE ADMINISTERED: 89/02/06 INSTRUCTIONS TO CANDIDATE: Una separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing-grade requires at least 70%'in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts. l % OF l l CATEGORY .% OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 24.94 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM) 27.00 26.93 2. EMERGENCY AND ABNORMAL PLANT. EVOLUTIONS (27%)' 1 1 48.25 48.13 ~3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC RESPONSIBILITIES (10%) Y/i T-140-2 % TOTALS FINAL GRADE All work done on this examination is my own. I have neither given nor received aid. Candidate's Signature J4XSTER COP Ee#-.c COPY
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS , During the administration of this examination the following rules apply: 1 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties. l
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil only to facilitate legible reproductions. I 1
- 4. Print your name in the blank provided on the cover sheet of the l examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers.
1
- 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category __" as i appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer sheet.
- 9. Number each answer as to category and number, for example, 1.4, 6.3.
- 10. Skip at least three lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
1 l 12. Use abbreviations only if they are commonly used in facility literature. l i 13. The point value for each question is indicated in parentheses after the ! question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the exami..dtion are not clear as to intent, ask questions of the examiner only.
i 17.'You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in i completing the examination. This must be done after the examination has ' been completed. k___---___--__________ _.____.m_ . _ _ _ _ _ . _ _ _ _ . __:._____.__________
18.-When you~ complete your examinat' ion, you shall:
- a. . Assemble.your examination as follows:
(1) Exam questions on top. (2) Exam aids - figures, tables, etc. (3) Answer pages including' figures which are part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
- d. Leave the examination area, as defined by the examiner. Iffafter leaving, you are found in this area while the examination is still.
in progress, your license may be denied or revoked. l l l l l
r - EQUATION SHEET f = ma v = s/t v = mg Cycle efficiency = Het Work (ou:*, - s = v,t + hat 2 Energy (in) E = mC a = (vg - y )/t KE = bmv v g=v + at A = AN A = A,e PE = agh w = e/t 1 = in 2/tg = 0.693/tg , W = v4P AE = 931am t g(eff) = (t,,)(ts) _ (tg+t)b U2=$ CAT ,' - IX
, P , I = I Q = UAAT o
, , P ur = W ' In I=Ie-VX g ,
g I=1 9 to-X/ M P=P 10 SUR(t) TVL a 1.3/u P=P o et/T EVL ' O.693/u
~SUR = 26.06/T ~
T = 1.44 DT - SCR = S/(1 - K,gg)
/1 p sUR = 26 'f f )1 g,p CR x = S/(1 - K,ff )
T = '(1*/o ) + [(i 'p)/A,,fp ] CR 1 (1 - Keff)I = CR2 (I ~ xeff)2 '~ T ,= t*/ (, ;-) M = 1/(1 - K,gg) = CR g/CR 0 T = G - p)/ A,gg p 9"I eff -IIIKeff " AEeff/K,ff g
- CI ~ Eeff)0 II ~ Eeff)1 p= ~
(1 /TK,"gg3 + @/(1 + A,gfT )) , 1* = 1 x 10 seconds P = E(V/(3 x 1010) A,gg = 0.1 seconds
-I E = Na 1d33=1d2y WATER PARAMETERS Idg =Id2 1 gal. = 8.345 lbm 2 R/hr = (0.5 CE)/d (meters)
I gal. = 3.78 liters 3 R/hr = 6 CE/d (feet) 1 ft = 7.48 gal. MISCELLANEOUS CONVERSIONS . Density = 62.4 lbm/ft 3 1 Curie = 3.7 x 1010dps Density = 1 gm/cm 3 1 kg = 2.21 lba Heat of valorization = 970 Etu/lbm I hp = 2.54 x 103 BTU /hr Heat of fusica = 144 Btu /lba 1 Hw = 3.41 x 100 Btu /hr 1 Atm = 14,7 psi = 29.9 in. I'g. 1 Btu = 778 f t-lbf 1 ft. HO-O,4335lbf/in2 g. inch = 2.54 cm 2 F = 9/5 C + 32 "C = 3/9 (*F - 32)
- 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS- Page. 4 (7%)-AND COMPONENTS (11%) (FUNDAMENTALS EXAM) i QUESTION 1.01 (0.75)
The reactor power is on IRM range 7 and the reactor is above the-Point"of Adding Heat (POAH). ' Figure 1.1 shows the Integrated Rod Worth (IRW) curves for a center control rod.. What is the shape-of the differential rod ' worth curve for the center control rod? (Sketch curve) (0.75) 1 l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) ;
( l l 1 delta k/k inch of rod movement CENTRAL CONTROL R0D O e ** FIGURE 1.1 l
L1. -REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 5
-(7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM) i QUESTION- 1.02 (1.25)
A reactor startup is in progress. The reactor is critical'and below the- Point of ' Adding Heat (POAH), reactor period is
' infinity, and IRM's are on range 1. At time t=0, a control rod is withdrawn and at time t=1, control rod motion is stopped.
Figure 1.2 'shows a plot .of Keff during the control rod withdrawal. What is the shape of the reactor period response curve from time t= 0- to time t= 1+7 (Draw response on figure 1.2, shape only) (1.25) I i
'(
l I 1 l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
i i l I I l l , g t l 8 j I I 4 ..
.I I .
Reactor Period ao - -- - -f---,---------- i l I i
, I i
i ! \ l I I l
, I A I l I i
I i l
, i .co i ~_ _ -.1 _ -
, i Keff Io
.--- - - -t - - - - - -
I l _ i ; I I I I I i m 0-o i n+ TIMC FIGURE 1.2
- 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS- Page 6 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION 1.03 (1.00) I j MUL11PLE CHOICE (Select the correct answer.)- l-The fission process produces PROMPT and DELAYED neutrons. Which of the following defines a DELAYED neutron?
- a. A high energy neutron born later than 10^-14 seconds after fission from the radioactive decay of various fission products.
- b. A neutron with an energy level of less than 0.025 Mev born before'10"-04 seconds following a nuclear fission.
- c. A neutron. emitted as a result of a gamma-neutron or an I
alpha-neutron reaction.
- d. A neutron emitted in less than 10"-14 seconds following~ a nuclear fission.
l l l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
m- _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ -_ 1 I
- 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS POgo 7 ]
(7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM) j 1 QUESTION 1.04 (1.00) i MULTIPLE CHOICE (Select the correct answer.) Which of the following statements is correct concerning the concentratiori of Xenon in the reactor core?
- a. Following a reactor scram from 100% reactor power, a xenon free condition will be reached in 20-25 hours. j
- b. Following a startup to 100% reactor power, Xenon concentration will reach an equilibrium value in approximately 4-7 hours.
- c. Following a reactor scram from 100% reactor power, peak xenon conditions will be reached in 7-11 hours.
- d. Following a power decrease from 100% reactor power to 50%
reactor power, xenon concentration will initially decrease
- and then increase to a new higher steady state value.
1 I 1 l l J (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
- 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 8 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
. QUESTION 1.05 (1.00)
MULTIPLE CHOICE (Select.the correct answer.) Control. rods, burnable poisons, and fission product poisons all affect the reactivity in the core. Which ONE of the following defines reactivity?
- a. Keff=1
- b. The fractional change in neutron population from one generation to the next.
- c. Ratio of neutrons produced from fission to those lost due to-leakage and resonance absorption.
- d. Delta k/k/ degree F.
l l l l l l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
=______-_-____-_____-_ _ _ _ _ _ .
a;
- 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 9 (7%) AND COMPONENTS-(11%) (FUNDAMENTALS EXAM)
- QUESTION 1.07 (1.00)
MULTIPLE CHOICE (Select the correct answer.) Which of the following best. represents the magnitude of the value. for the void' coefficient at BOL (beginning of. life)? Void (delta k/k/% void)
- a. 1 x 10^-03
- b. -1 x 10^-05
- c. -1 x 10"-03
- d. -1 x 10^+03 J
l i i i
.]
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) i 1
_ _ - _ - - - _ _ _ _ _ _ - _ - _ _ _ - - _ _ _ _ _ - - . _ - - _ = . i
'1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Paga 10 )
(7%) AND COMPONENTS (11%) ' (FUNDAMENTALS - EXAM) QUESTION . 1.08 (1. 00) . MULTIPLE CHOICE ' (Select the correct answer. ) i Non-condensible gases affect the operation of the condenser. j j l Which ONE of the following describes the effect on the plant of an. INCREASE in Non-condensible gases in the. main condenser?
- a. Condenser pressure decreases '(vacuum. increases)
- b. Circulating-water outlet temperature decreases, i'
- c. Steam cycle efficiency decreases.
- d. Condensate depression increases.
I d 1 \ l i l l i I (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) Am-__u_m___m____. ___m________-____e.m_ _ _ _ . _ _ _ . a.- _ --- ___
1
- 1. REACTOR ~ PRINCIPLES (7%) THERMODYNAMICS- Page 11 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM) 1 i
QUESTION 1.09 (1.00) MULTIPLE CHOICE (Select the correct answer.)
.Which of the following statements correctly describes the limiting condition for the LHGR thermal limit?
- a. Clad temperature of 2200 degrees F. !
- b. 1% plastic strain on the clad.
- c. Onset of transition boiling (OTB) around the cladding. 1
- d. Total. pin power exceeds 13.4 KW. ,
l 1 1 1 l 1 1 I l l 1 1 (***** CATEGORY 1 CONTINUED-ON NEXT PAGE *****) J l u _ _ _ ___ _
- 1. REACTOR' PRINCIPLES (7%) THERMODYNAMICS .Page'12 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION 1.10 (1.00) MULTIPLE CHOICE (Select the correct answer.) Which one of the following statements correctly describes the operation of a SJAE? (Refer to the attached diagram 1.4 of a steam jet air ejector.)
- a. The. steam jet entering the suction chamber mechanically.
entrains the non-condensible condenser gases.
- b. The highest velocity of the steam and air mixture occurs at the throat of the air ejector.
- c. The highest velocity of the steam flow entering the suction chamber is independent of the pressure (vacuum) in the main condenser.
- d. The. highest steam pressure occurs at the throat of the steam jet during normal operation.
I 1 1 i l l i l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) I
m l l
\
l l
- STEAM N0ZZLE 2- Z AIR EJECTOR - r-s1m"
=
//7 suc110~ N / 01sceAR2 l
CHAMBER THROAT -- j
/
/ 1
/
s l 1 FROM CONDENSER 1 1 P
"'e #*
ese e 'Ndame e e FIGURE 1.4
1.- REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 13 (7%) AND COMPONENTS 111%) (FUNDAMENTALS EXAM) QUESTION 1.11 (1.00). 1 MULTIPLE CHOICE (Select the correct answer.) ! Which of the following correctly describes an: operational characteristic of a jet pump? (See figure 1.5)
- a. The diffuser portion of the jet' pump' decreases the velocity of the flow leaving the throat (mixing section) of the jet-pump.
- b. The suction' flow rate through the jet pump must be greater j than the driving flowrate. '
i
- c. The highest static pressure experienced.in the jet pump is _
at the exit of the jet pump driving nozzle. q
- d. The static pressure at the exit of the jet pump diffuser is f equal to the pressure at the entrance to the driving nozzle. j i
l l l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) t
._ ___ __- ______=_ a
i ORIVlHG FLOW t DRIVING N0ZZLE 4 SUCTION FLOW t II iI l ( THROAT OR MIXING SECTION li k f -l j .. f s v
/
( i 5 DIFFUSER i i Y Jet' Pump Operating Principles FIGURE 1.5 j
y--- - _ . -
~1. REACTOR PRINCIPLES (7%) THERMODYNAMICS -Page 14.
(7%) AND COMPONENTS (11%) (FUNDAMENTALS' EXAM) QUESTION 1.12 (1.00) MULTIPLE CHOICE (Select'the correct answer.) Which ONE.(1) of the followin'g statements correctly describes the
'offect of fast neutrons on the reactor _pressue vessel?
- a. Increase the Ductility.
- b. Decreases the stress which must be applied to the RPV to cause plastic deformation.
- c. Increases probability of brittle fracture.
- d. Decreases crack growth rate.
1 , l l l l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) L---___-__________----___
- 1. ' REACTOR PRINCIPLES-- (7%) - THERMODYNAMICS Page 15 (7%);AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION. 1.13 (1.00) Multiple Choice (Select the correct answer.) Which of the following selections defines NPSH?
- a. . Combination'of static'and dynamic pressure on the suction of the pump. -
- b. Fluid energy' lost due to friction.
.c. Head resulting strictly from fluid motion.
- d. Difference between total pressure and' saturation pressure.
(***** CATEGORY' 1 CONTINUED ON NEXT PAGE *****).
- _ - _ _ - _ _ _ - _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ = _ _ _ .. - _ _ _ _ -_ __________=_____-_________-______-____
- 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 16 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION 1.14 (1.00) MULTIPLE CllOICE (Select the correct answer.) The Reactor Building is normally maintained at a negative pressure of 0.25 inches of water to help minimize the spread of radioactive contamination. If atmospheric pressure outside of the Reactor Building is 14.8 psia, what is the absolute pressure inside the Reactor Building?
- a. 14.55 psia
- b. .123 psia
- c. .01 psia
- d. 14.79 psia
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) i
- 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 17 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION 1.15 (1.00) The attached figure 1.3 shows a basic closed fluid system with its pump characteristic curve. The two pumps are IDENTICAL, variable speed, radial centrifugal pumps. An operator closes down on throttle valve V-10, but does not shut the valve completely. How would this effect the system operating curve? (Draw shape on Figure 1.3) (1.0) b (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
m I
"'O TalA E" f 7 &3 Nr v'- i (ofdn) .b V- 7 flosed)
X7, w$,lHie) - N bPneM " f'"" E g%,m 9acc a aacc e 7" f A ' T7 V-+ V-G
. n (open) , @losed)
I
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V-)O ,' VN TH RCTTLEt> i Y-V si M IN VN (cfen ) fu,,.p 6 ' Gaven Sysbe m 1 StArV8 org& I)U m I) A T YCAk (n/N4/71HM Sf[O 1 i i F~ lo u.> CLOSEDFLUIDSYSTEMSCHEMATICWITHONEPUMPANDSYSTEMCubE FIGURE 1.3
- 1. ' REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 18 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
I QUESTION 1.16 (1.00)
. MULTIPLE CHOICE - (Select the correct -answer)
Ju reactor. recirculation pump is limited.to two starts'in succession'when the motor windings are initially at ambient temperature to prevent damaging the pump motor. l Which of the following describes the cause of-the~ motor damage?-
- a. Counter emf exerts excessive torque on the pump shaft.
- b. Excessive stator current causes the rotor to accelerate to rated speed too quickly.
- c. Excessive armature current may cause overheating of- the windings.
- d. Heating of motor windings prevents counter emf from building up.
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
- 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 19 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION 1.17 (1.00) MULTIPLE CHOICE (Select the correct answer. ) The reference leg on the "A" narrow range level detector, which is providing the control signal to the Feed Water Level Control-(FWLC), ruptures. Which of the following correctly describes the response of the reactor plant?
- a. Feedwater flow rate decreases.
l l b. Feedwater flow rate increases,
- c. Actual reactor water level increases.
- d. Indicated level, for the sensor with the failed reference leg decreases.
e I l l 1 l 1 I I l
]
l l l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) 1
- 1. REACTOR' PRINCIPLES (7%) THERMODYNAMICS Page 20
.(7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION 1.18 (1.00) MULTIPLE CHOICE (Select the correct answer.)- The Reactor Vessel Instrumentation narrow range. and wide range reactor water level instruments have a condensing. chamber atop the reference leg. What is the purpose of the condensing chamber?
- a. To prevent reference leg flashing during rapid depressurization of the reactor vessel.
- b. To alleviate the need for density compensation by keeping the reference leg temperature close to the temperature of the variable leg.
- c. To keep the reference leg filled by condensing steam.
l d. To provide pressure compensation for the reactor pressure I exerted on the variable leg. (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
l
- 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 21 ;
(7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM) I l
)
I
. QUESTION 1.19 (1.00) l MULTIPLE CHOICE (Select the correct answer) i 1
A centrifugal pump (i.e., ECCS pump) should not be operated with its discharge valve shut. Which of the following selections is the reason?
- a. Prevent overspending the pump.
- b. Prevent internal pump damage,
- c. Prevent overstressing valve disk,
- d. Prevents electrically overloading pump motor.
l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) l
I 1.- REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 22 (7%) AND COMPONENTS (11%) -(FUNDAMENTALS EXAM) QUESTION 1.20 (1.00) MULTIPLE CHOICE (Select the correct answer.) { The Source Range Monitor (SRM) detectors operate in the proportional region of the Gas conductivity curve while the Intermediate Range Monitors (IRM) and Local Power Range Monitors-(LPRM) operata in the ionization region.l(See F.igure 1.6) Which ONE selection listed below describes the reason for operating the SRM's at higher voltages than the IRM's or LPRM's?
- a. Greater sensitivity to neutrons.
- b. Prolongs detector 31fe.
- c. Ensures gamma radiation response is proportional to neutron level.
- d. Prevents ionization of Argon gas.
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
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)
)
I D LIMITED ' g.y PROPORTIONAL PLATEAU, REGION, 'l f SEGION F
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. . . _ . . . , , g CONTINUOUS l
l DISCHARGE F REGION Z W 2 C e D g PROPORTIONAL U REGION l g lONIZATION l l O REGION 0-o A l(k % W RECOMBINATION REGION l 8 " I e _ O l J l l
! l I 0
I i vI y 2 v 3 v y, ] 4 APPLIED VOLTAGE l l l l I Regions of Detector Operating Characteristics Gas Conductivity Curve FIGURE 1.6 , i
- 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 23 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION 1.21 '(1.00) . MULTIPLE CHOICE (Select the-correct answer.) A centrifugal pump is. operating at 1800 RPM with its motor drawing 4 amps. The pump speed is then doubled to 3600 RPM. (Assume voltage remains constant). Which selection represents the motor current at 3600 RPM?
- a. 1.58 amps
- b. 2 amps
- c. 32 amps
- d. 64 amps
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
- 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Pago 24 ,
(7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM) ! i
)
QUESTION 1.22 (1.00) J l l There are Four (4) reasons for using the Reactor Water Cleanup and Condensate Domineralizers to maintain reactor water chemistry within specification. What are TWO (2) of the FOUR (4) reasons? (0.5 each) (1.0) j l I l I { l l l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
-1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 25 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION 1.23 (1.00) ONE (1) of the_four (4) sources of non-condensible gases in the main condenser is air in-leakage. What are TWO (2) of the remaining THREE (3) sources? (0.5'each) (1.0) l l
]
l i l (***** CATEGORY '1 CONTINUED ON NEXT PAGE *****)
'1. ' REACTOR PRINCIPLES-(7%) THERMODYNAMICS Page'26 (7%) AND COMPONENTS (11%)-(FUNDAMENTALS EXAM)
QUESTION- 1.24 (1.00) MULTIPLE CHOICE (Select the correct answer.) Fluid hammer is a term used for an overall pressure transient on a. system. Which of the following selections is the cause of fluid hammer? a.- Slowly closing a valve.
- b. Rapidly changing the flow rate in a piping system.
- c. Mechanical agitation.
- d. Excessive fluid viscosity.
l l- (*****-CATEGORY l CONTINUED ON NEXT PAGE *****) l _=-____________-_-____:_____-_-_________-____-________. -_ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - - _ _ _ _ - - - _ - - _ _ _ _ _ _ _ _ _ - -
h
- 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Pago'27 )
(7%) AND COMPONENTS (11%) ( FUNDAMENTALS EXAM) i . QUESTION 1.25 (1.00) The attached figure 1.7 illustrates the pump characteristic' curve for one (1) of two (2) identical, single speed centrifugal pumps which are to operated in parallel. The operator starts both pumps. What does the combined pump characteristic curve look like? (Draw) (0.25 for pressure head, 0.5 for shape, 0.25 for flow)
-(0.75)
)
(***** END OF CATEGORY 1 *****) 1 l l
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**--***one- .ww.,,
s FIGURE 1.7
- 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 28 (27%) 1 I
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'2 . ' EMERGENCY 'AND ABNORMAL PLANT EVOLUTIONS Page 29 (27%)
I 1 l QUESTION 2.01 (1.75) i The. reactor is operating at 20% rated thermal power when reactor water i level begins rapidly increasing. Answer the.following questions in { accordance with QOA 201-8, High or High High Reactor Water Level. i
- a. Why is the operator directed to maintain reactor water level below l
+60 inches? (0.50) q
- b. What are FIVE (5) COMPONENTS-(SYSTEMS) that may suffer major. damage due l
to excessively high vessel level during power operation? (1.25) 1 i l l l J; i l' (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
I 1
- 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 30 (27M l
i QUESTION 2.02 (1.75) 1 The reactor is operating at 85% of rated power with the Recirculation
~
f System in Master Manual Control when Recirculation Pump A speed rapidly j
. increases to 104%, but a reactor scram does NOT occur. l Answer the following questions concerning-QOA 202-2, Reactor Recirculation l.
System Failure - Flow Controller Fails High.
- c. What is the Technical Specification BASIS for the immediate operator action requiring reactor recirculation pump speeds to be within 10% of each other? (0.75)
I
- b. What are TWO-(2) IMMEDIATE ACTIONS that the operator is directed to ]
perform if he is unable to adjust recirculation pump speeds to within ' 10% of each other? (1.00) l i 1 (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) l L--_____________________
- 2. EMERGENCY'AND ABNORMAL PLANT EVOLUTIONS Page 31 (27%)
QUESTION 2.03 (2.00) The. reactor is operating at 90% rated power when a loss of feedwater heating results in a 50 deg. F decrease in the feedwater injection temperature. Answer the following questions concerning QOA-400-1, Reactivity Addition, and QOA 3500-1, Loss of Feedwater Heaters, immediate operator actions.
- a. What is the BASIS for the immediate operator action which directs the operator to reduce power by reducing recirculation flow PRIOR TO inserting control rods? (0.50)
- b. How is the operator directed to INSERT control rods?
(FOUR (4) STEPS REQUIRED.) (INCLUDE in your discussion ANY Protective Features which must be overcome and ANY verifications which must be made.) (1.00)
- c. When should control rod insertion be stopped? (0.50) l l
(***** CATEGORY. 2 CONTINUED ON NEXT PAGE *****)
m - ;
,2. ' EMERGENCY AND-ABNORMAL PLANT EVOLUTIONS -Page 32 'l (27%) l b
f QUESTION 2.04-A'small LOCA has caused a reactor scram. Tractor water level is l . steady at 50 inches and reactor p ,essiire Js 800 psig, . decreasing slowly (5 psig/ minute),'and Core Spray hakA utontatically initiated due to a.high drywell pressure.
'f i
l- .How flow valve will the Core Sp h)oop (MO-1402-g8B Bl,s bs stem fails.tC ED? respond ifa the (INCLUDE associated' description minimum. of the= Core Spray loop B flowpat in your)<11 ussion.) (Ther IS'Wo (1.00) o/L e 4 % . >
-1
)
.i l
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(***** CATEGORY- 2 CONTINUED ON NEXT PAGE *****) l i
~c 1 2 .- EMERGENCY AND ABNORMAL PLANT EVOLUTIONS page 33 (27%)
)
QUESTION 2.05 (1.50) i j The reactor is operating at~100% rated power when an off-gas explosion- 1 occurs causing the. radioactivity release rate of Technical Specification j 3.8.A.to be exceeded. ' Answer the following questions concerning QOA 5400-2, Off-Gas Explosion - Recombiner Operating and Filters Bypassed. !
- a. What is the REASON for~the immediate action which directs the operator to SLOWLY reduce load at 20-MWe/hr in order to reduce the' radioactive
' release below the Technical Specification limit?
(TWO (2) REASONS REQUIRED.) (1.00)
- b. The immediate actions direct operators to verify the Off-Gas system integrity following the explosion. What RADIOLOGICAL CONTROLS should operators exercise when approaching the steam jet air ejector rooms?
(0.50) I ~
)
J
-l i 1
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) l l
j 4
- 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 34 i (27%) a QUESTION 2.06' (2.50)
The'ResidualfHeat Removal System has initiated in'the LPCI mode due to a . j small LOCA resulting in a High.Drywell pressure. Reactor pressure is 700 psig and decreasing very slowly. Answer the.following in accordance with QOA_1000-4,.LPCI Automatic Initiation. ) a.- If RHR loop B minimum flow valve fails closed, what are.the FOUR (4) IMMEDIATE ACTION STEPS which must be performed to establish a flow path j for RHR loop B. (2.00) l O l b. What ACTION must the operator take if no' flow path can be established l for RHR Loop B. (0.50) ] \ l l l l l 1 i I l l (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
~
l
- 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 35 (27%)
QUESTION 2.07 (3.00) Answer the following questions concerning the Emergency Operating Procedures.
- a. What is the BASIS for General Caution number 3, which directs the operator to determine Suppression Pool temperature by averaging the channel A and Channel B temperatures of recorder 1640-9 on panel 901-21. (0.75)
- b. General Caution number 4 provides the following information for determining whether the reactor water level instruments may be used for vessel level indication.
Drywell Temperature Indication Level Reactor water Level Instrument 361 deg F -309 inches GEMAC Lower 400 Range (-334 to +66 inches) _ What CONDITION (S) must be met (using the Table above) to prevent use of the GEMAC Lower 400 Range water level instrument for level indication? (0.50)
- c. General caution number 9 prov.4 des ':he following conversion chart for determining actual vessel level using the GEMAC Lower 400 Range.
Reactor Pressure Actual Vessel Indicated Level (Inches) Level (inches) 500 psig Rx. Press 1000 psig Rx. Press
+30 -40 -66
+10 -57 -80
-60 -113 -130
-140 (~ Top of Cere) -177 -188
-240 -257 -260
- 1. Why is the GEMAC Lower 400 Range indicated level less than the actual vessel water level at reactor pressures of 500 psig and 1000 psig? (0.75)
- 2. What are the TWO (2) CONDITIONS which must be met in order to use this conversion chart? (1.00)
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
- 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 36 (27%)
l 1
- l l
QUESTION 2.08 (3.00) { i l Answer the following questions concerning QGA 100-3, RPV. Power. control.
- a. When is the operator required to enter QGA 100-3, RPV Power Control?-
(0.50)
- b. What is the BASIS for step B-3, which directs the operator to " Confirm or initiate a recirc pump runback to minimum speed" prior to: tripping
.both' recirculation pumps in step C-3? (INCLUDE ADVERSE CONSEQUENCE OF TRIPPING RECIRCULATION PUMPS WITHOUT REDUCTION IN SPEED. ) (1.00)
- c. Why does step M-1 direct the operator to "Close 301-25 (CRD charging isolation valve". (0.50)
- d. Why does step H-5 direct the operator to " Inhibit ADS by placing the auto blowdown inhibit switch _to INHIBIT"? (0.50)
- e. Why_do steps F-5 and H-5 direct the operator to." Inject SBLC" if the reactor CANNOT be shutdown before Suppression Pool. temperature exceeds 110 deg. F? (0.50) l l
l I i (***** CATEGORY 2-CONTINUED ON NEXT PAGE *****) L___ _____ _ _ _ _ _ _ _ _ _ _ _ _ . _ ______
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- 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 37 (27%)
QUESTION 2.09 (1.00) MULTIPLE CHOICE (Select the correct answer.) Select the ONE statement which explains why step D-3 of QGA 200-3, Primary Containment Pressure Control, PREVENTS the operator'from initiating Suppression Chamber sprays when Suppression Chamber pressure is ABOVE 12.5 psig.
- a. To prevent excessive thermal shock to the Suppression Chamber spray ;
ring. l
- b. To prevent depressurizing the Suppression Chamber so rapidly that the negative design pressure of the Suppression Chamber is exceeded.
- c. To prevent raising the level of the Suppression Chamber above the Drywell to Suppression Chamber vacuum' breakers.
- d. To prevent chugging at the downcomer discharge.
l l 1 1 1 (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)-
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- 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 38 (27%)
QUESTION 2.10 (2.00) Unit 2 reactor plant is operating at 75% rated power when a loss of the number 2 125V DC power supply occurs (i.e. number 2 125V Battery is lost). Answer the following in accordance with QOA 6900-4, Loss of Unit 2 125V DC Supply.
- a. MULTIPLE CHOICE (Select the correct answer.)
Select the ONE SIGNAL (METHOD) which will trip the Unit 2 Main Turbine. (1.00)
- 1. Depressing the turbine trip pushbutton in the control room
- 2. A high vibration trip signal
- 3. Operating the local turbine trip lever at the turbine front standard
- 4. A Main Generator trip
- b. MULTIPLE CHOICE (Select the correct answer.)
Which ONE of the following Unit 2 systems will NOT automatically function during the loss of 125V DC power. (1.00)
- 1. ADS (Auto Depressurization System)
- 2. Unit 2 Emergencf Diesel Generator
- 3. HPCI (High Pressure Coolant Injection)
- 4. RCIC (Reactor Core Isolation Cooling)
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
- 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 39 (27%)
QUESTION 2.11 (1.00) MULTIPLE CHOICE (Select the correct answer.) A loss of Shutdown Cooling has occurred on Unit 1. The reactor has been shutdown for the past 12 hours, both recirculation pumps have-been shutdown, the head vents have been opened, and shutdown cooling had been in service for one hour. Which ONE of the following plant conditions would be indicated when the operator observes the upper vessel area metal temperatures increasing. (ASSUME NO OPERATOR ACTION IS.TAKEN.) ;
- a. Proper natural circulation is occurring 1
i
- b. The core is being cooled by reverse-flow through the core
- c. Adequate core cooling is assured by steam cooling through the head vents ,
l
- d. Temperature stratification of the vessel coolant is occurring !
t } I l l (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) i
. .j t________._____________.___________.___________._._______.__. . . _ _ . _ _ _ _ _ _ _ _ . _ . . . _ . _ .
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- 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 40 (27%)
QUESTION 2.12 (2.50) A Unit 1 reactor startup is in progress and reactor power is 10% of rated when a leak in the Unit 1 Instrument Air header causes the Instrument Air pressure to STEADILY DECREASE.
- a. SELECT the ONE correct answer for each of the blanks ((1) through (3)) i from the corresponding column listed below for the following statement.
The low instrument air LOW pressure alarm will be received when air pressure has decreased to (1) psig, the service air backup stop ("Little Joe") valve will open when instrument air pressure decreases to (2) psig and the instrument air dryer bypass valve will open l at (3) psig. (0.50 each) (1.50) ; COLUMN (1) (2) (3) 1 , l (a) 100 (a) 105 (a) 110 i (b) 87 (b) 95 (b) 100 (c) 77 (c) 85 (c) 90 , (d) 70 (d) 75 (d) 80 )
- b. MULTIPLE CHOICE (Select the correct answer) j Which one of the following valves will fail CLOSED on a COMPLETE LOSS l of Instrument Air. j
- 1. CRD flow control valve (1.00)
- 2. Feedwater regulating Low flow control valve
- 3. Reactor Building to Torus Vacuum Breakers (1601-20A and -20B)
- 4. Reactor Feed Pump recirculation (minimum flow) valve l
l l (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
.2. -EMERGENCY AND ABNORMAL' PLANT EVOLUTIONS Page 41 (27%) QUESTION 2.13 (1.00) 46ULM&t B-1DMOICE (Solo L uno mviimmo . . . _ _ _ , _ A malfunction of the feedwater regulating valves has forced the operators to control reactor water level by alternate means as outlined in QOA 600-1, Feedwater Regulating Valve Lockup. State the'TWO (2) m METHODS of controlling reactor water level per this procedure, if reactor water 1 vel is increasing. (0.5 each) (1.0) (G,w - % G 4' d 1%~ , l 1 l l i i (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) l
2.- EMERGENCY AND ABNORMAL PLANT EVOLUTIONS .Page 42 (27%) QUESTION 2.14 (1.50) !
-)
i l 1 A fuel element failure causes-a SUSTAINED high-high radiation signal from the radiation elements in the "30 minutes holdup volume". QOA 1700-4 requires four'(4) .immediate actions, one of which is to verify sustained ; high-high radiation level on the Off-Gas radiation monitors. l What'are the THREE (3) remaining immediate actions? (0.5 each) (1.50)
~
l l l \ l l . l
)
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) i j
)
- 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Pago 43 (27%)
QUESTION 2.15 (1.50) l During refueling operations, a fuel bundle is dropped into the reactor i vessel. Bubbles are then seen rising in the water from the vicinity of the 1 dropped fuel bundle. One (1) of the immediate actions listed in QOA 800-1, l Irradiate Fuel Damage While Refueling, is to refer to QGA 300-1. What are the THREE remaining immediate actions? (0.5 each) (1.50) i 1
)
(***** END OF CATEGORY 2 *****)
- 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 44 RESPONSIBILITIES (10%)
-l I 1
l I l i i 1 l l i j l
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- 3. ' PLANT SYSTEMS (38%)'AND PLANT-WIDE GENERIC Page'45 j RESPONSIBILITIES (10%) i j
QUESTION 3.01 (2.50)- ] As a result:of an Anticipated Transient Without. Scram (ATWS), it has become-necessary to initiate Standby Liquid Control (SBLC). l l a. What are four (4) of theffive (5) positive reactivityLcontributions which must be overcome by the SBLC system? (0.25 each)' (1.0) q
=l
- b. Given' the : positions of SBLC system initiation switch aus being " system -
2&l", " system 2", "off", " system 1", and'" system 1&2", how does the l l operator position this switch to initiate SBLC. (0.5)
.1
- c. What are four (4) of the seven (7) parameters checked to verify SBLC I initiation?-(0.25 each) (1.0) i
- ;l l
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i 1 i 1 (***** CATEGORY' 3 CONTINUED ON NEXT-PAGE *****) l 1 1 _. ____ _ __i;
m.
- 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GEMERIC Page.46 RESPONSIBILITIES (10%)
. QUESTION 3.02. .(1.00)
MULTIPLE CHOICE (Select the correct answer.) { A reactor startup is in progress and power has just come on scale on the ; Intermediate Range Monitors (IRM's). What power should be indicating on the' Source Range Monitors?'
- a. 10 - 10^2 cps..
- b. .5 x.10^2 - 10^3. cps. j
- c. 10^4 - 10^5 cps,
- d. 6 x 10^5 - 10^6 cps.
1 l' (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) J __.n_____________.__________.___. _ _ _ _ _ _ . . _ _ . _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ - . - ____._________-_________.______________.______._____l
-3.- ' PLANT SYSTEMS-(38%) AND PLANT-WIDE GENERIC Page'47 RESPONSIBILITIES (10%)
QUESTION 3.03 (1. 00) MULTIPLE CHOICE (Select the correct answer.) A reactor startup is in progress and cower indicates 10'3 cps on the' Source-Range Monitors (SRM's). A malfunction occurs such that +\- 24 VDC power - from the batteries to SRM A is interrupted. What is the impact of this malfunction?
- a. SRM A Drive mechanism Motor Module inoperative.
- b. Ionization of Argon gas in SRM A ceases.
)
- c. SRM A "short period" alarm.
i
- d. SRM A inoperable. i 1
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-3.- PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 48 RESPONSIBILITIES (10%) QUESTION 3.04 (1.00)
' MULTIPLE CHOICE (Select the correct answer.)
SRM's provide rod blocks to aid in preventing the operator from adding ex-cessive reactivity to the core during a startup. Which ONE (1) of the conditions below will cause:all SRM rod blocks to be
-bypassed?
a.- All IRM's on range 3 or above.
- b. SRM's' read greater than 100 counts.
- c. All IRM's are on range 8 or above.
- d. SRM's read greater than 2 X 10^5 counts.
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_ _ - = _ - _ - . _____ _-__. --_____-___-___-___-___ -___-_-___ -_-_____ _ -
J3 . - PLANT SYSTEMS (38%)~AND PLANT-WIDE GENERIC' Page 49 RESPONSIBILITIES (10%) QUESTION- 3.05 (1.00)
~
The Reactor Manual Control System controls two (2) sets of valves in the Control. Rod Drive Hydraulic (CRDH) system. What are these two sets of valves? (0.50 each) l l l l 1 (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) l 1 .
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3.- PLANT SYSTEMS ~(38%) AND PLANT-WIDE GENERIC Page 50 RESPONSIBILITIES (10%) , QUESTION 3.06 (2.75) The High Pressure Coolant Injection-(HPCI) system automatically initiates-on low-low reactor water level.(-59").following a.small instrument line break. Reactor water level reaches +48" causing a HPCI turbine. trip just as.a high HPCI area temperature alarm (200' degrees F) causes a system isolation.
- a. WhatLis the other HPCI automatic INITIATION signal? (include setpoints)
(0.25-initiation- 0.25 setpoint). , (0.5)
- b. What are the remaining two (2) signals which will cause a Group IV system ISOLATION? '(include-setpoints) (0.25 for signal, 0.25 for setpoint). (1.0)
- c. What are the three (3) DIRECT automatic actions which occur on-a.HPCI j system isolation? (0.25 each) (0.75)- 'I
- d. How is the HPCI system isolation reset? (0.5)
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q 1 3.- PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page'51 RESPONSIBILITIES (10%) J l o i i QUESTION- 3.07 (1.50) ( I The reactor is operating at 100% rated power. Two'(2) narrow range l pressure compensated GE\MAC's indicate' reactor water level from 0"Lto 60". , A malfunction occurs such that the GE\MAC's'are no longer pressure i I compensated. Reactor pressure increases and ACTUAL water level remains constant.
- a. What happens to indicated GE\MAC water level? (0.5)
- b. Why are all reactor water level trip signals obtained from the Yarway instead of GE\MAC instruments? _( 0.5)
- c. What is the purpose for.the auxiliary chamber on the Yarway instruments? (see figure 3.1)- (0.5) j 4
1 i 1
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3 . -- PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 52 RESPONSIBILITIES (10%) QUESTION 3.08 (1.00) ] MULTIPLE CHOICE (Select the correct answer.) A small LOCA has. occurred, causing the Emergency Diesel Generators (EDG)-to c auto start on high drywell pressure. EDG #1 trips shortly after startup. Which of the following would cause #1 EDG to trip under these conditions?
- a. High. jacket water temperature.
- b. Engine overspeed.
- c. Low lube oil temperature.
- d. Generator overvoltage.
}
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- 3. PLANT SYSTEMS (38%)'AND PLANT-WIDE GENERIC Page 53 RES POSSIBILITIES (10%)
QUESTION 3.09 (1.00) MULTIPLE CHOICE (Select'the correct answer.) Drywell pressure of +2.5 psig causes the Standby Gas Treatment (SBGT) System-to autostart. Which of the following will also cause the SBGT system to autostart?
- a. Main Steam line high radiation (3X normal).
- b. Reactor Building Ventilation Exhaust Radiation (3 mrem /hr.)
- c. Reactor high pressure. (1060 psig.)
- d. Reactor building high pressure. (+0.25 in) 1 l
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1
' 3 .- PLANT SYSTEMS-(38%) AND PLANT-WIDE GENERIC Page 54 RESPONSIBILITIES (10%)-
l QUESTION 3.10 (1.25) l l Figure 3.2 is a block diagram of the Standby Gas Treatment ( SBGT) .. system. 1 The names of several of the components have been omitted. l
- a. What are the following components identified in the.figuro? (0.25 each)
Component 1? Component 27 Component 3?
- b. What is the purpose of the electric heater? (0.5) 1 l
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- 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 55 RESPONSIBILITIES (10%)
QUESTION 3.11 (1.50) When charging the' Control Rod Drive Hydraulic system accumulators with nitrogen, per QOP 300-6, Control Rod Drive System Accumulator Charging, .the QOP 300-T1-accumulator pre-charge pressure is determined from figure s73
- a. Why is the accumulator pre-charge pressure HIGHER (-&%E psig) at an ' ambient temperature of 90 deg. F than the pre-aharge pressure (552 psig) at 50'deg. F? (0.50)
- b. What ADVERSE CONDITION would result from pre-charging an accumlator to a pressure of 375 psig at an ambient temperature of 70-deg. F?-
(0.50)
- c. What is the reason for NOT EXCEEDING the pre-charge pressure as ;
determined by figure QOP 300-T1. (0.50) , i i i I l ! 1 i
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1
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v- - QOP 300-T1
.Ravision 1 September 1983 ACCUMULATOR PRE-CNARGE NITROGEN PRESSURE VERSUS AMBIENT TEMPERATURE 120 i....,i . . . .- . . -
i- i 1 110 - 100 -
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, ACCUMULATOR PRE-CHARGE NITROGEN PRESSURE VERSUS AMBIENT TEMPERATURE (final) , u v a..
COT 0 61983 1 0.c. u. s. n. l - - - - - - _ - _ _ - - - _ - _ _
l
- 3. PLANT' SYSTEMS (38%) AND' PLANT-WIDE GENERIC Paga 56 l RESPONSIBILITIES (10%)
i QUESTION 3.12 (3.00) l a)- How do the Backup Scram Valves respond'to a trip on.RPS Channel A? ) (Both valves open, Backup scram valve A opens, Backup scram valve B opens, neither Backup Scram valve opens.) (0.5) l b) What is the power supply to the Backup Scram Valve soleniods? (0.5) j 1 c) What is the response of the following valves if Backup Scram Valve B is inadvertently opened during testing? (OPEN, CLOSE, REMAIN OPEN, OR j REMAIN CLOSED) J i
- 1. Scram pilot solenoid valves. .(0.5) i
- 2. Inboard Scram Discharge Volume drain valves. (0.5)
- 3. Scram inlet valves. (0.5)
- 4. ATWS scram solenoid valves (0.5) l
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- 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 57
_ RESPONSIBILITIES (10%) QUESTION 3.13 (1.00) , MULTIPLE. CHOICE (Select the correct answer.) A LOCA occurs due to a small break.in an instrument line outside the drywell. Reactor water level decreases to -59" and remains there for 8.5 minutes causing the Automatic Pressure Relief System.(APRS) to automatically initiate'and start all low pressure ECCS pumps. The' Shift Engineer decides to terminate APRS operation. Which of the following describes how this may be done?
- a. Place the automatic blowdown inhibit keylock switch tx) inhibit.
- b. Place.Drywell Pressure Reset keylock switch in RESET.
- c. Turn off Residual Heat Removal pumps.
- d. Turn off Core Spray pumps.
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- 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 58 RESPONSIBILITIES (10%)
QUESTION 3.14 (1.00) MULTIPLE CHOICE (Select the correct answer.) The APRS logic circuit will automatically initiate APRS when specific key plant criteria have been achieved. What is the power supply to the APRS logic?
- a. RPS bus A\B.
- b. Instrument bus.
- c. 24/48 VDC system.
- d. 125 VDC turbine building bus.
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1 3? PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC 'Page 59 RESPONSIBILITIES (10%) , 4 i l l . QUESTION 3.15 (1.00) MULTIPLE CHOICE (Select the correct answer.) ~I-
' il A-reactor startup is'in progress and power is less than 10%. -The Nuclear ) -Station. Operator (NSO) is withdrawing a control rod when a " SELECT BLOCK" l is received.
3 What is a possible cause for this alarm? ;
- a. CRD withdrawn.past position "48"
- b. High~ water level in Scram Discharge Volume.
- c. Timer malfunction. 1 i
- d. APRM downscale (3/125 full scale).
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.3. PLANT SYSTEMS (38%) AND PLANT-WIDE' GENERIC Page 60 l RESPONSIBILITIES (10%) j i
QUESTION 3.16 (1.00) An Anticipated Transient Without Scram (ATWS) is in progress and the. reactor operator attempts to insert the control rods with the " Emergency In" feature. This overrides all rod blocks with the exception of two.(2). .
- a. What are two (2) blocks associated with the Reactor Manual Control.
System that would prevent insertion of rods with the " Emergency In" , feature? (0.5) I
- b. How is the circuitry for_the " Emergency In" feature different from that of the " Continuous In" feature? (0.5) l i
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- 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC 'Page 61 RESPONSIBILITIES (10%)
~ QUESTION 3.17 (2.75)
The plant is operating at 64% power as indicated by APRM's and the 'j recirculation system is providing 50% of rated flow through the core. The i reactor operator selects a centrally-positioned control rod for withdrawal. The Rod Block Monitoring System (RBM) determines local power level to be j
'62%, but a malfunction in the RBM gain change circuit prevents increasing the local power level signal to the reference value of 64%.
- a. What is the effect of the malfunctioning gain circuit on the Rod Block 1 Monitoring System? (0.5)
- b. What will cause the RBM to shift from the normal to alternate reference I
APRM? -(0. 5)
- c. What are the two (2) conditions under which the RBM is AUTOMATICALLY bypassed? (0.5 each) (1.0)
- d. What are the three (3) purposes of the Rod Block Monitoring System? I (0.25 each). (0.75)
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- 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC- Page 62 g RESPONSIBILITIES (10%) i
. QUESTION 3.18 (2.75) ;
I 1 The reactor operator intends to increase recirculation pump speed from twenty (20) percent to fifty (50) percent; however the speed limiter in the Recirculation Flow Control system will limit pump motor speed to 30% unless the pump' discharge valve is open and feedwater flow is greater than twenty percent (20%) of rated. l a. What are the power supplies to the reactor recirculation pump drive motors? (0.25 each) (0.5)
- b. What is the reason for limiting pump speed with the discharge valve shut? (0.5)
- c. What is the reason for limiting pump speed with with less than 20%
feed flow? (0.5)
- d. How does the reactor recirculation system respond to an Anticipated a Transient Without Scram? (include ALL applicable trip-signals, setpoints, and recirculation system response (s) ; 0.25 each) (1.25) [
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- 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 63 RESPONSIBILITIES (10%)
QUESTION 3.19 (2.75) The Reactor Water Cleanup System is in the Blowdown Mode to control reactor water level during a reactor startup when an Anticipated Transient Without Scram (ATWS) occurs. The operator initiates Standby Liquid Control and one (1) of the four (4) ways the RWCU system responds is to shut the inboard suction / isolation valve.
- a. What are three (3) other responses of the RWCU system? (0.75)
- b. What two (2) other conditions will cause the RWCU system to isolate?
(include initiation and setpoint) (0.5 each) (1.0)
- c. How will the Blowdown Flow Control Valve respond (open fully, throttle open, close fully, or remain as is) following the initiation of SBLC?
Explain your answer. (1.0) (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) l
- 3. PLANT SYSTEMS-(38%) AND PLANT-WIDE GENERIC Page:64' RESPONSIBILITIES (10%)
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QUESTION- 3.20 (2.50) ( 1 Answer the following. questions concerning the Off Gas system. i
- a. Why does the Pre-heater use steam instead of electric heaters? (0.5)
- b. What is the purpose of the charcoal beds? (0.5)
- c. What radiation level in the main steam lines.will cause a system isolation? _(0.25)
- d. What are the five.(5) COMPONENT actuations on a main steam line high radiation associated with'the Off Gas system? (0.25 for each component-actuation.) (1.25) l
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- 3. PLANT ' SYSTEMS (38%)' AND PLANT-WIDE GENERIC Page 65 RESPONSIBILITIES (10%)
QUESTION 3.21 (1.00) MULTIPLE CHOICE (Select the. correct answer.) The reactor is operating at 80% power. The Drywell radiation monitoring syste is operating normally. The green OPERATE lamp on the readout module is illuminated and the radiation level meter is reading normally. What is the normal radiation level for the Drywell monitor?
- a. 1 mrem /hr.
- b. 500 mrem /hr.
- c. 1 R/hr.
- d. 10 R/hr.
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- 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 66 RESPONSIBILITIES (10%)
QUESTION 3.22 (1.00) MULTIPLE CHOICE (Select the correct answer). The reactor is operating at 80% power with the following feed and condensate pump alignment: Feed pumps A and B running. Condensate Booster Pumps A, B, and C running with condensate booster pump D in standby. l Condensate pumps A, B, and C running. ] When will the standby condensate pump start?
- a. When feed pump suction pressure is less than 160 psig.
- b. When condensate flow decreases below 95% full flow.
- c. When the A condensate pump breaker opens.
- d. When hotwell level increases to 33 inches.
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- 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC- - Page 67 RESPONSIBILITIES (10%)
-]
QUESTION. 3.23 (1.50) i A reactor startup is in progress. The operator is withdrawing control rods j and observes the lights on the Full Core Display. ! 1 What is indicated by each of the following lights on the Full Core Display? (Two-conditions required for each, 0.25 each.)
-a. A BLUE LIGHT for rod 34-27 (0.5)
- b. An accumulator trouble light for rod 16-25. (0.5)
- c. A red rod drift light for rod 18-23 (0.5) 1 i
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- 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Pago 68 RESPONSIBILITIES (10%)
QUESTION 3.24 (1.00) MULTIPLE CHOICE (Select the correct answer). The amount of two (2) phase flow resistance within a fuel bundle increases as power increases. This has a tendency to re-route flow to the lower power bundles. Specific design features have been incorporated to minimize the impact of increased flow resistance due to two (2) phase flow. Which selection represents one (1) of these design features?
- a. Placing lower enriched fuel at the periphery of the core.
l b. Flow orifices in the fuel support pieces.
- c. Minimizing length of active fuel region.
- d. Limit Linear Heat Generation Rate (LHGR) to 13.4 KW/FT.
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- 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 69 RESPONSIBILITIES (10%)
L -QUESTION 3.25 (1.00) Techn'ical Specifications' limit reactor pressure when irradiated fuel'is in the vessel. l c. What is this reactor pressure limit? (0,5)
- b. Where'is this reactor _ pressure measured? (0.5) l
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- 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Pago 70 RESPONSIBILITIES (10%)
QUESTION 3.26 (1.00) l MULTIPLE CHOICE (Select the correct answer.) The standby condensate pump is in the pull to lock position due to , maintenance in progress. An ORANGE ring is around the pump control switch. ! This ring designates that the equipment is temporarily out-of-service. ! What is the maximum length of time the ORANGE ring may be used?
- a. 2 hours.
- b. 4 hours.
- c. 6 hours.
- d. 8 hours.
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'3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC. Page 71 RESPONSIBILITIES (10%)
QUESTION. 3.27 (1.00) A reque.st to place condensate pump "A" out-of-service has been initiated. The Station Control Room-Engineer (SCRE)' has completed his review.of the ' task and passes the Master Out-of-Service checklist to the center desk Nuclear Station Operator (NSO) for processing. When the NSO completes processing the Master Out of Service checklist: Who does the NSO pass the Checklist on to?
- a. SCRE
- b. Communications Center Coordinator.
.c. Shift. Engineer.
- d. Maintenance Worker.
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- 3. -PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 72 RESPONSIBILITIES (10%)
QUESTION 3.28 (1.50) Maintenance must be performed in a controlled area in which the general area. radiation levels are 40 mrem /hr. 'The Radiation-Chemistry Supervisor has authorized whole bcdy doses in excess of the daily maximum dose written in the Radiation Protection Standards, QRP 1000-1. The operators radiation history is as follows: Weekly dose 0 mrem Quarterly dose 1000. mrem-Annual dose 3.6 Rem I Lifetime dose 60 Rem Quarterly extremities dose 1100 mrem Quarterly skin dose 1100 mrem Radiation History (NRC Form 4) On File Age 40 years
- a. What is the operators most restrictive exposure limit? (0.5) i b. What is the maximum length of time the operator can stay in the area?
l (0.5 for application, 0.5 for value) (1.0)
**SHOW ALL WORK **
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- 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC 'Page 73 RESPONSIBILITIES-(10%)
QUESTION 3.29 (1.00) ; MULTIPLE CHOICE (Select the correct. answer.) Which ONE of the following' cards-is used to HOLD equipment'out-of-service for the protection of personnel working .cn1 equipment.
- a. Personnel Protection card
- b. Caution card
- c. Hold' card
- d. Out-of-Service card l
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,3. PLANT SYSTEMS (3 8%)' AND PLANT-WIDE GENERIC Page 74 .
RESPONSIBILITIES (10%) 1 QUESTION 3.30 (1.00), MULTIPLE CHOICE (Select the correct answer.). The Atmospheric Containment dilution (ACAD) system helps maintain the hydrogen concentration below the lower explosive limit. What is;the lower explosive limit for hydrogen?
- a. 2%
l b. 4%
- c. 6%
- d. 96%
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j 3.' PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 75 RESPONSIBILITIES (10%) QUESTION 3.31 (2.00) A fire in the turbine building is reported to the control room over the Public Address system. The Nuclear Station Operator (NSO) is required to take four (4) actions per QAP 1170-3, Fires. What are the four (4) actions? (0.5 each) (2.0) I 9 (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
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- 3. : PLANT SYSTEMS- (38%) AND PLANT-WIDE GENERIC Page-76 RESPONSIBILITIES (10%)
QUESTION- 3.32 (1.00) MULTIPLE CHOICE (Select'the correct answer.) l Under certain circumstances, it is permissible to perform only one (1) ] ! valve position verification of a safety related~ system during during.an { p operating valve lineup check. Which selection represents one (1) of these circumstances?
~
)
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- a. When authorized by the Nuclear Station Operator.
] 'b. When remote indication is available.
- c. In cases which involve significant radiation exposure.
- d. When the valve is listed on the Out-of-Service sheet.
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- 3. PLANT SYSTEMS-(38%) AND PLANT-WIDE GENERIC Page 77 RESPONSIBILITIES-(10%)
QUESTION 3.33 (1.00) MULTIPLE CHOICE (Select the correct answer.) While walking through-the turbine building, an operator comes across a sign that reads: CAUTION RADIATION AREA What is the maximum dose rate an individual may be exposed to in this area?
.a. 5 mrem /hr.
.b. 99 mrem /hr.
- c. 110 mrem /hr.
- d. 100 Rem /hr.
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l '. . REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 78 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
-ANSWER 1.01 (0.75) see attached sheet i
1 1 REFERENCE i Quad Cities Reactor Theory Text, page 58. Quad Cities Learning Objective, Reactor Theory, #10. 1 292005K107' ..(KA's) I q ANSWER 1.02 (1.25) .] l see attached sheet REFERENCE Quad Cit'ies, Reactor Theory, pages 38-46. Quad Cities, Learning Objectives, Reactor Theory, #7. 292008K112 ..(KA's) ANSWER 1.03 (1.00) a. REFERENCE Quad Cities, Reactor Theory, page 38. . 1 Quad Cities, Learning Objectives, Reactor Theory, #8. 292001K102 ..(KA's) 1 l l l (***** . CATEGORY 1 CONTINUED ON NEXT PAGE *****)
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- 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 79 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
ANSWER 1.04 (1.00) c. REFERENCE Quad Cities, Reactor Theory, page 70. Quad Cities Reactor Theory Lea.ining Objectives, #11. 292006K107 ..(KA's) ANSWER 1.05 (1.00) b. REFERENCE
- Quad Cities Reactor Theory, pages 26-28.
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) C (.1 ' d (* .
REFEREN Quad Cities Rea xt pages 32-34. Quad Cities Learning Objectives, eactpr Theory, #6. 292003K101 ..(KA's) ~ ANSWER 1.07 (1.00) c. (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
- 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 80 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
REFERENCE Quad Cities Lesson Plan, Reactor Theory, page 54 Quad Cities Learning Objectives, (not covered) l 292004K114 ..(KA's) ' i ENSWERs 1.08 (1.00) 1
'~ \,N 'Tkit' k e'
. j s j REFERENCE
' ~s,' 1
s ,
General Electric Heat Transfer and Fluid Flow, page 4-Co. J Ocad Cities Learning Objectives, (not covered). I 293007K107 ..(KA's) 'N. I l . ANSWER 1.09 (1.00) b. REFERENCE General Electric Heat Transfer and Fluid Flow, page 9-16. Quad Cities, Learning Objectives, HTFF, #18. 293009K107 ..(KA's) ANSWER 1.10 (1.00) a. REFERENCE General Electric Heat Transfer and Fluid Flow, page 4-32. Quad Cities, Learning Objectives, HTFF, 4. 293004K104 ..(KA's) 1 l l l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
- 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 81 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
ANSWER 1.11 (1. 0 0) a. REFERENCE ] Quad Cities HTFF, page 8. ( Quad Cities, Learning Objectives, HTFF, 4. j 293004K105 ..(KA's) l ANSWER 1.12 (1. 00) c. REFERENCE General Electric BWR Academic Series on Material Science (NOV 84) ) page 6-16. I Quad Cities Learning Objectives, (not covered) 293010K105 ..(KA's) - ANSWER 1.13 (1. 00) , 1 d. REFERENCE General Electric Heat Transfer and Fluid Flow, page 6-114. Quad Cities HTFF, page 6. Quad Cities, Learning Objectives, HTFF, #6. 293006K110 ..(KA's) (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
- 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 82 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
ANSWER 1.14 (1.00) d. REFERENCE General Electric Heat Transfer and Fluid Flow, page I9 to I-10. Quad Cities Primary / Secondary Containment, pages 4,.46. Calculation Sheet 293001K101 ..(KA's) ANSWER 1.15 (1.00) l see attached curve. (1.0) l REFERENCE General Electric Heat Transfer and Fluid Flow, page 6-102. l Quad Cities Learning Objective, HTFF, #6. 3 291004K108 ..(KA's) l l l I ANSWER 1.16 (1.00) c. REFERENCE l GE BWR Academic Series, Electrical Science, page 6-39. j Quad Cities Learning Objectives, (not covered). l 291005K106 ..(KA's) (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
_- -~ -' V l
- -Hm+ Le Etckcsv.er I**Y
' 7tl'3 h V- 5' (ofd"h b V- 7 (closed)
. ](U'ci# led N ,fy b onen4 F b <"e4
/
(,pe,) _ Cooler A Cooler B _ Pu w p A T7 V-4 b V-6 A bee-) ^ (cInsed) Wlo b FotTlab v-2 st tv1 IN VN b/on) eu...e e a ,, c Sysie m,
,qu s9 TEM OPCi2.AT144 cv W E dur Vd.
O N E. I)LL m I3 l-] T Nead puiwmum sp gD N s N
\
/ \
,.<*.../' \
F lo u.> CLOSED FLUID SYSTEM SCHEMATIC WITH ONE PUMP AND SYSTEM SURVE FIGURE 1.3 - KEY
- 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 83 (7%) AND. COMPONENTS (11%) (FUNDAMENTALS EXAM)
ANSWER 1.17 (1.00) a. REFERENCE Quad Citics, Vessel Level Control System. ; Quad Cities, Vessel Level Control System Learning Objectives, #8. l 291002K109 ..(KA's) ' l ANSWER 1.18 (1.00) c.
- Reference General Electric Academic Series on Instruments and Controls, i l
page 2-30 Quad Cities Reactor Vessel Instrumentation, Figure 1. Quad Cities Learning Objectives, Reactor Vessel Instrumentation, #1. REFERENCE
-291002K107 ..(KA's)
ANSWER 1.19 (1.00) b. REFERENCE GE Heat Transfer and Fluid Flow, pg 4-53 Quad Cities Learning Objectives, (not covered). 291004K104 ..(KA's) ANSWER 1.20 (1.00) a. (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
- 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 84 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
REFERENCE Quad-Cities Source Range Monitor System, pages 4-5. Quad Cities Source Range Monitor System Learning Objective #7. 291002K119 ..(KA's) ) ANSWER 1.21 (1.00) l c. REFERENCE
~1 Quad Cities HTFF, page 6. !
Quad Cities HTFF Learning Objectives, #6. - 291005K104 ..(KA's) -1 l ANSWER 1.22 (1.00) { (2 required, 0.5 each)
- 1. Minimize impurity deposits (on core surfaces).
- 2. Limit presence of impurities (which enhance corrosion of primary system).
- 3. To minmize the concentration of corrosion products.in the primary coolant.
1
- 4. To maintain water clarity sufficient to perform refueling operations.
~ REFERENCE ; 1 Quad Cities BWR Chemistry, page 2. I Quad Cities BWR Chemistry Learning Objectives, #1. I 291007K103 ..(KA's) j (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) 1----.- --- ----------u_---
- 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 85 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAh, ANSWER 1.23 (1.00)
.(2 required, 0.5 each)
- 1. Activation gases from the irradiation of reactor coolant.
- 2. Fission gases (produced in the fuel will' leak to the coolant through cladding defects).
- 3. Radiolytic decomposition.
REFERENCE Quad Cities, Off Gas System, page 20. ( Quad Cities Off Gas System Learning Objectives, #4. 291006K118 ..(KA's) ANSWER 1.24 (1.00, l b. l REFERENCE General Electric Heat Transfer and Fluid Flow, page 6-56. Quad Cities Learning Objectives, (not covered). 291006K114 ..(KA's) ANSWER 1.25 (1.00) see attached Figure 1.7-KEY l REFERENCE General Electric Heat Transfer and Fluid Flow, page 7-116 to 7-117. l 291004K113 ..(KA's) (***** END OF CATEGORY 1 *****)
)
+
9 l I i h (FT" cF WATE32.) 2.0 - I l 40 12.o i
\
AN NM 4 FIGURE 1.7-KEY
-2. EMEF.GENCY AND ABNORMAL PLANT EVOLUTIONS Page 86 (2711 ANSWER 2.01 (1.75)
- a. To maintain level witnin'the limits of his indication on panel'901-5 (902-5). (0.50)
- b. 1. HPCI turbine y CJx.> c M W 5" O '
- 4. Main steam line piping
- 5. Reactor vessel
- 6. SRVs (5 required, 0.25 ea.)
REFERENCE QOA 201-8, PG 1 - 3. 295008A101(3.7/3.7) 295008K101(3.0/3.2) 295008K205(3.8/3.9) 295008K206(3.4/3.6) 295008K208(3.4/3.5) 295008K211(3.1/3.3)
...K/A (K/A VALUE) 295008K208 295008K206 295008K205 295008K101 295008A101
..(KA's)
ANSWER 2.02 (1.75)
- a. To prevent the LPCI Loop Select logic from selecting the wrong' loop for injection (in the event of a LOCA). (0.75)
- b. (2 of the following required, at 0.5 each)
- 1. Trip Recirc pump A
- 2. Place the controller for recirc pump B in manual
- 3. Reduce load (pump speed to prevent recirc pump B overload 724 AMP) .
MCA ct-e m 2c Q S, cm uq_A" %A % C; c; A 2.cs L~ 4 * (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
-2. EMERGENCY AND ABNORMAL' PLANT EVOLUTIONS Page 87-(27%) REFERENCE QOA 202-2 PG 1 AND QOA 202-4 PG 1. TECH SPEC BASES PG 3.6/4.6-24. 295001A101(3.5/3.6) 295001A1059(3.3/3.3) 295001G010(3.8/3.7)
-295001K205(3.2/3.6) ...K/A (K/A VALUE) 295001K205 295001G010 295001A105 295001A101 ..(KA's)
ANSWER 2.03 (2.00) ( m y % m - a Ac J J m -(>sm a u Y 4 w A f c 4 4 te A , ,u a-pewW% mq.
- a. APRMflowbiaslineh)asbeenselectedtoallowadequatemarginto avoid a scram as flow is reduced. (reduc'ed from 0.65w to 0.58w) c.J ac 'Pc w ~
- W c.t~.~ 4 0.50)
- b. 1. Place RWM mode switch to bypass "g,4 W ' , 2.5lPO- (cudd)
- 2. Insert control rods in reverse sequence
- 3. Continuously insert to full in (any rod inserted)
- 4. Check the flow contro! line after each group of rods inserted (4 required, 0.25 ea.)
- c. When the reactor is operating below the 100% flow control line. (0.50)
REFERENCE QOA 400-1 PG 1 AND 3. QOA 3500-1 PG 2. 295014A104(3.2/3.3) 295014A204(4.1/4.4) 295014G007(3.3/3.6) 295014K104(3.0/3.4) 295014K209(3.4/3.6) 295014K211(3.6/3.7)
...K/A (K/A VALUE) 295014K209 295014K104 295014G007 295014A204 295014A104
. . ' ( KA ' s )
ANSWER 2.04 ( '
. 00)h ) '
The core spray loop B discharge re ef v ' e (RV-1402-28B) will lift and divert water to the RBEDT from e to'us i6also accept appropriate discussion of this transfer water.) / (1. 0 0) - b Y '
" d*WM i
lf} /
,f h' j)2 -
Mb9 l f' ili} k d4/ W W K N O S d"" d'! (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) 3b7ff. _ - _ _ - 1
l
- 2. EMERGENCY'AND ABNORMAL PLANT EVOLUTIONS Page 88 )
(27%) l REFERENCE
- QOA 1400-1 PG,2.
295036A203(3.4/3.8) 295036K201(3.1/3.2) ...K/A (K/A VALUE) 295036K201 295036A203 ..(KA's) ANSWER 2.05 (1.50) l
- a. The slow power change will minimize the increased release of long-lived l
fission' products. .(0.50) and l Will also minimize the increase in the release rate'of radioactive l iodine (from the enimney and the reactor building vent' stack). (0.50) j
- b. Full protective clothing (0.50) (and full face mask)
(NA: ww & acybs/4 ) l REFERENCE l QOA 5400-2 PG 1, 2, AND'4. 295038A101(3.9/4.2) 295038A204(4.1/4.5) 295038G007(3.2/3.5) 295038K102(4.2/4.4) 295038K302(3.9/4.2) ...K/A.(K/A VALUE) , l 295038K302 295038K102 295038G007 295038A204 295038A101
..(KA's)
ANSWER 2.06 (2.50) L a. 1. Place the CONTAINMENT COOLING PERMISSIVE (S-17) switch to ON (0.50)
- 2. Place the CONT. CLG. 2/3 LEVEL & ECCS INIT BYPASS (S-18) keylock switch to MANUAL OVERRIDE (0.50)
- 3. Open MO 1001-34B (Suppression Pool Test Return Isolation Valve)
(0.50)
- 4. Throttle open'MO 1001-3 6A/ B (Suppression Pool Test Return Valve)
(0.25) to establish 12,00 gpm.,, (0.25) ; CL +1 s . f i x ut. w . :.u, J h _l
- b. Place RHR Loop B pumps in PULL-TO-LOCK. (0.50)
REFERENCE QOA 1000-4 PG 1. 295031A101(4.4/4.4) 295031G010(4.0/3.8) ...K/A (K/A_VALUE) . 295031G010 295031A101 ..(KA's) j (***** CATEGORY 2 CONTINUED ON NEXT PAGE'*****)
2.: EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 89
-(27%)
ANSWER 2.07 (3.00)
- a. The suppression pool is a very large volume where wide variations in temperature may occur, resulting. in local: temperatures which. may not be representative of the entire volume. (Actions within the emergency operating procedures ~(QGAs) are based upon bulk (average).
suppression pool temperature.) (0.75)
- b. Drywell temperature exceeds 361 deg. F-(0.25) and.the. instrument reads below -309 inches. (0.25) l C. 1. GE\MAC lower 400 is calibrated cold (0 psig, 70 degF) (0.75)
- 2. Recirculation pumps off (0.50)
Drywell temperature (approximately)~ normal (0.50) REFERENCE LESSON PLAN ON EMERGENCY OPERATING PROCEDURES PG'12, 14, AND 15. l LEARNING OBJECTIVE NO. 4. 295026A103(3.9/3.9) 295026A201(4.1/4.2) 295026G007(3.4/3.8) 295028A203(3.7/3.9) 295028G007(3.4/3.8) 295028K101(3.5/3.7) 295028K203(3.6/3.8) ...K/A (K/A VALUE) 295028G007 295028A203 295026G007 295026A201 295026A103
..(KA's) l l
h (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
v
'2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 90 (27%)
ANSWER 2.08 (3.00)
- a. ' Anytime'that QGA 100-1,-RPV Level Control, is entered. (0.50)
(Also accept: listing.ALL entry conditions for QGA 100-1.)
- b. Provides for more control of the RPV level transient (0.50). Without the speed reduction'a turbine trip (+48 inches) may occur due to the level swell (0.50).
- c. To prevent' diverting flow (pressure) from the drive. water header'to allow maximum CRD pump discharge for driving control rods manually.
'(0. 5 0)
(Also accept: To allow FCV to open and supply flow for driving control l- rods' manually.) l i d. To prevent a cold. water transient due to automatic injection of systems during depressurization. (0.50)
- e. To assure a reactor shutdown will be achieved before the HCTL of the suppression pool is exceeded. (0.50) 1 1
REFERENCE LESSON PLAN ON EMERGENCY OPERATING PROCEDURES PG 42 - 45. LEARNING OBJECTIVE NO. 6.
,295037A206(4.0/4.1) 295037G007(3.7/3.9) 295037G011(4.4/4.7) 295037K204(4.4/4.5) 295037K205(4.0/4.1) 295037K301(4.1/4.2) 295037K302(4.3/4.5) ...K/A (K/A VALUE) 295037K205 295037K204 295037G011 '295037G007 295037A206
..(KA's) 1 ANSWER 2.09 (1.00) d.
l l l ' l l 1 (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) l
.y
.2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 91 (27%)- REFERENCE LESSON PLAN ON EMERGENCY OPERATING PROCEDURES PG 53. LEARNING OBJECTIVE NO. 6. 295024G007(3.6/3.9) 295024K213(3.8/3.8) 295024K302(3.5/3.8)
...K/A (K/A VALUE) 295024K302 295024K213 295024G007 ..(KA's)
-4 ANSWER- 2.10 (2.00)
- a. 3. (1.00)
- b. 4. (1.00)
REFERENCE LESSON PLAN ON LEARNING OBJECTIVE NO. QGA 6900-4 PG 1, 2, AND 5. 295004A102(3.8/4.1) 295004G007(3.0/3.5) 295004K203(3.3/3.3)
...K/A (K/A VALUE) 295004K203 295004G007 295004A102 ..(KA's)
ANSWER 2.11 (1.00) d. REFERENCE LESSON PLAN ON j
. LEARNING OBJECTIVE NO. J QOA 1000-2 PG 4. ;
295021A205(3.4/3.5) 295021K102(3.3/3.4) 295021K103(3.9/3.9) 295021K104(3.6/3.7) ...K/A (K/A VALUE) 295021K104 295021K103 295021K102 295021A205 ..(KA's)
)
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) i
- 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 92 (27%)
ANSWER 2.12 (2.50) 1
- a. (1) (b) 87 (2) (c) 85 (3) (d) 80 (3 required, 0.50 ea.)
- b. 1. (1.00) ,
1 REFERENCE LESSON PLAN ON CRD HYDRAULICS PG 22 LESSON PLAN ON FEED AND CONDENSATE PG 26. LEARNING OBJECTIVE NO. 4. QOA 4700-2 PG 1. 295019A101(3.5/3.3) 295019K201(3.8/3.9) 295019K203(3.2/3.2) 295019K301(3.3/3.4) ...K/A (K/A VALUE) 295019K301 295019K203 295019K201 295019A101 ..(KA's) ANSWER 2.13 (1.00) l (0.50)
- 1. Reject water with RWCU system
- 2. (If necessary) maintain reactor water level by adjusting reactor power. (0.50) l%( CM $.Yck M ca d 4 L ' o C c M " ^ - <3 ^ -J c A ~
REFERENCE ph vm c c QOA 600-1, Feedwater Regulating Valve Lockup. 295009A101 295009K202 ..(KA's) i i i (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
2 .- EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 93 (27%) i ANSWER 2.14 . ( 1. 5 0 ) (3 required, 0.5 each) Run recirculation pumps down to minimum speed. Manually SCRAM the. reactor. Prior to the 15 minute timer expiring, shut AO-1-5408A & B-(AO-2-5408)
. 2 T occ yAJ 10 hb / l:.s, > A. ll t 9 .~ e* -
REFERENCE QOA 1700-4, Off Gas High Radiation Monitoring. 295017K203 295017A102 ..(KA's) ANSWER 2.15 (1.50) (Three required, 0.5 each) (1.50) (Use the public address system) to evacuate the reactor building. Verify initiation of the Standby Gas Treatment system. Verify isolation of the reactor building ventilation. REFERENCE QOA 800-1, Irradiated Fuel Damage While Refueling. Q-u-_---,-_---,----------- - - _ - - - - - . - - - - - - . - - - , - - - - - - - - - - - - - - - - - - - - - - - - - - - ' - - - - -
295023K101 295023G010 ..(KA's) 1
-1 I
i 1 1 i I 1 I 1 l l l 1 (***** END OF CATEGORY 2 *****) l l 1 I l l l l i i L_._.-._____._____._._.__ _ _ _ _ _ _ - .
3.--PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 94 RESPONSIBILITIES (10%) , i ANSWER 3.01 (2.50)
- a. (Four required, 0.25 each)
Elimination of steam voids. 1 Moderator temperature change ~(from hot to 125 degrees F.) Reduced doppler. Decreased control rod worth as the moderator cools. Xenon decay.
- Shutdown margin 3% dK. l Reactivity associated with control rods not. inserted.
l
- b. Place the switch to the system 1&2 position (through positon system 1) )
(Also accept: to the system 2&1 position (through system 2 position) ' j (0.50)
- c. (Four required, 0.25 each)
Flow indicating pilot light lit Decreasing SBLC tank level SBLC pump discharge pressure > reactor pressure. Neutron flux level decreasing. Reactor Water Cleanup System isolates. Squib continuity light goes out. Squib valve circuit failure annunciator. 1 i REFERENCE QOP 1100-2 Quad Cities Standby Liquid. control, page 10. Technical Specification, 3.3 . 211000A305 211000A302 211000A301 211000A408 211000K503'
..(KA's) i ANSWER 3.02 (1.00) c.
REFERENCE . Quad Cities Intermediate Range Monitors, page 17.
. Quad Cities Intermediate Range Monitor Learning Objectives, #9.
215003A301 ..(KA's) (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
.3. 1 PLANT SYSTEMS-(38%)'AND PLANT-WIDE GENERIC Page 95 RESPONSIBILITIES (10%) j ANSWER 3.03 (1.00) 1
- d. ]
I REFERENCE l
. Quad Cities Source Range Monitors, figure 1 l
Quad Cities Source Range Monitors Learning Objectives, #4. 215004G007 215004K201 215004K602 ..(KA's)
)
ANSWER 3.04 (1.00) t
- c. !
REFERENCE Quad Cities Source Range Monitors, pages 13, 14. Quad Cities Source Range Monitors Learning Objectives, #5. 215004K401 215004K102 ..(KA's) I ANSWER 3.05 (1.00) : W i Stabilizing valves. Directional control valves.
"v Whd
^
^ '
(0.50)
'( 0. 50) in.
i . REFERENCE Quad Cities Control Rod Drive Hydraulic system, pages 14, 28, 58. Quad Cities Control Rod Drive Hydraulic Learning Objectives #2, #13. 201002K102 ..(KA's) 1
-(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
m -
- 3. PI ANT SYSTEMS (38%) AND PLANT-WIDE GENERIC ~Page 96 RESPONSIBILITIES (10%)
ANSWER 3.06 (2.75)
- a. (0.25 each)-
High.drywell pressure 2.5 psig (+\-'.25psig) _ (0.5)
- b. (0.25 each: .
3 Low reactor. pressure 100 psig (+\- 10 psig) (1.0). High steam line flow 300% (+\- 30%)
- c. (0.25 each) . (0.75)
Auxiliary oil pump interlocked against autostart. Suppression pool suction line isolates, m cu _9 . sum a i Steam Supply.line isolates, g. 3 3 3w ,.h u-t;. m j
- d. Depress the " Reset valves 4+5" pushbutton. m %id hu. 6M O . 5) t
& 4 - yd " "
w . REFERENCE Quad Cities High Pressure Core Injection, page 4, 68. Quad Cities High Pressure Core Injection Learning Objectives,#5,#6,-#8. 206000K407 206000K404 206000K402 206000K116 ..(KA's) ANSWER 3.07 (1.50)
- a. Indicated level decreases (due to a lower water density seen by the dp cell which increases the d/p output between the variable leg and the reference leg.) (0.5) !
- b. Yarways require no power supply for local operation. -(0.5)
- c. Helps maintain the reference leg full during. rapid depressurization of the reacter vessel. (0.5) 1 REFERENCE !
Quad Cities Reactor Vessel Instrumentation, pages 2, 12, 14. Quad Cities Reactor Vessel Instrumentation Learning Objectives #1, #3. 216000K513- 216000K324 216000K301 216000K112 ..(KA's) . 1 l (***** CATEGORY l 3 CONTINUED W NEXT PAGE **t**) j l
m 3--. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 97 RESPONSIBILITIES (10%) ANSWER 3.08 (1.00) b. REFERENCE Quad Cities Diesel Generator, page 46. Quad Cities Diesel Generator Learning Objectives, #5. 264000K402 . . ( KA ' s ) - ANS 'ER 3.09 (1.00) b. REFERENCE l Quad Cities Standby Gas Treatment System, page 10. ) Quad Cities Standby Gas Treatment System, Learning Objective #9. I 261000A213 ..(KA's) l ANSWER 3.10 (1.25)
- a. (0.25 each) (0.75).
Component 1=Demister filter Component 2= Rough prefilter Component 3= Carbon Iodine Adsorber j
- b. Lowers the relative humidity (to allow more efficient operation of the charcoal adsorbers.) (0.5) j REFERENCE Quad Cities Standby Gas Treatment System, page 6, figure 1.
Quad Cities Standby Gas Treatment' System Learning Objectives #4, #5. 261000G007 ..(KA's) (***** CATEGORY 3-CONTINUED ON NEXT PAGE *****)
=_-____________________--____________ . _ - - - _ . - - _ _ . _ _ _ _ _
'3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 98 RESPONSIBILITIES (10%) ' ANSWER 3.11 (1.50) W cq " GU'~*W mew ge' W g .
c e--5"q=-*W W a., To prevent the accumulator nitrogen pressure from decreasing to the ' alarm setpoint when the ambient temperature-decreases. ( O '. 50 )
- b. The rod scram time will increase. (0.50)
- c. To ensure the required amount of water is available to scram the. drive. -
(0.50) REFERENCE l Quad Cities CRD Hydraulic System, page 14 and 16. Quad Cities CRD Hydruulic system Learning Objectives, #6, #11 201001A106 ..(KA's). ANSWER 3.12 (3.00) a) Neither Backup Scram Valve opens (0.5) ~! b) 125 VDC bus (0.5) c) 1. Remain closed ,4 es c% c_ , i g 4,A ,g
- 2. se o H e, _ . Ag o ,3 ,
- 4. Lamain closed * **" ' ~
(4 required, 0.5 each) REFERENCE I l LESSON PLAN ON 0500 RPS PF 8, 10, 12, 16. LEARNING OBJECTIVES NO. 3 AND 9. 201001G007 (3.6/3.7) 201001K203(3.5/3.6) 2 01001K4 04 ( 3. 6/3. 6) - l 201001K406(3.8/3.9) ...K/A.(K/A VALUE) 201001K406 201001K404 201001K203 201001G007 ..(KA's) i (***** CATEGORY 3 CONTINUED ON~NEXT PAGE *****)
- 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Pago 99 RESPONSIBILITIES (10%)
1 i ANSWER 3.13 (1.00) J l a. 1 REFERENCE l 1 Quad Cities Automatic Pressure Relief system, page 5. j Quad Cities Automatic Pressure Relief system Learning Objective #5. J 218000K403 ..(KA's) l ANSWER 3.14 (1.00) I i l l d. J REFERENCE I Quad Cities APRS page 11. l Quad Cities APRS Learning Objectives, (not covered). ! 218000K201 ..(KA's) ! a i q i ANSWER 3.15 (1.00) l c. l REFERENCE Quad Cities RMCS & RPIS, page 20. Quad Cities RMCS & RPIS Learning Objective 4. 201002K403 ..(KA's) (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
- 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page100 RES POSSIBILITIES (10%)
i ANSWER 3.16 (1.00) I
- a. (0.25 each) (0.5)
Rod Worth Minimizer insert block. Select blocks.
- b. " Emergency In" directly energizes the directional control valves (or )
bypass the Rod Drive timer). (The " continuous in" feature ~does not bypass the timer). c-c n.c<.g g cm (0.5) wu y% ' REFERENCE Quad Cities RMCS & RPIS, pages 10, 30. Quad Cities RMCS & RPIS Learning Objectives-#7, #9. 201002K105 201002K406 201002A402 ...(KA's) ANSWER ?.17 (2.75)
- a. A rod block signal is generated. (0,5)
- b. Placing the normal APRM in bypass (on the 90X-5 panel). (0.5) l l c. 1. Reference APRM < 30% power (0.5)
- 2. Edge rod selected (0.5)
- d. (0.25 each, 3 required) (0.75)
Prevent local' power from. approaching minimum critical power ratio limits during rod withdrawal. i Prevants local fuel damage (by supplementing the average power range i monitor rod block and scram trip functions). ) ' l Prevents gross overpower (in a local region from exceeding the total core power limits). REFERENCE l l Quad Cities Rod Block Monitor Sys eem, pages 2, 3, 5. ! Quad Cities Rod Block Monitor System Learning Objectives, #1, #2, #8. 215002K403 ..(KA's)' 215002K101 215002K301 215002G004 l l l i l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) l , 1 1 l .___- ________ - - - -
- 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page101-RESPONSIBILITIES-(10%)
4 ANSWER 3.18 (2.75) 1
- a. (0.25 each) .(0,5)
. Bus 11 Bus 12
- b. Excessive axial thrust could be developed across the pump. (0.5) i
- c. Prevents pump damage due to cavitation. (0.5) j i
I
- d. A high reactor pressure (0.25) of 1250 psig (0.25), energizes the trip coils for the recirculation. pump motor generator field breaker (0.25),- 1 l
reactor low water level (0.25) -59 inches (0.25) (1.25). REFERENCE J Quad Cities. Recirculation Flow Control System, pages 5. ] ~ Quad Cities Recirculation System, page 3. Quad Cities Anticipated Transient Without scram, page 8. Quad Cities ATWS Learning Objectives, #9. l 202001K201 202001K402 202001K506 ..(KA's) ANSWER 3.19 (2.75)
- a. (three required, 0.25 each) -(0.75) i Outboard suction / isolation valve shuts (MO-1201-5).
RWCU pumps trip. Flow demineralizers go into hold. RWCU return isolation / discharge valve shuts (MO-1201-80)
- b. (0.25 each) (1.0)
Low reactor water level Group III (+8",+/-2") , l NRHX high outlet temperature 140 degrees (+/- 5 degrees) j
- c. Close fully (0.25). The flow control valve will shut when pressure upstream of the FCV decreases to less than 5 psig (0.75). (1.0) i t
-(***** CATEGORY 3 CONTINUED ON NEXT,PAGE *****) ;
, ~ -
- 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page102 i RESPONSIBILITIES'(10%)
REFERENCE Quad Cities RWCU system, pages 4, 5, 6, 8, 10. Quad Cities RWCU Learning Objectives, #3, #5, #7. 204000A301 204000K404 ..(KA's) ANSWER 3.20 (2.50) i
- a. To eliminate an explosion hazard. (0.5) 1 i
- b. Delay _ noble gases (krypton and xenon)'by adsorbtion. (0.5)
I
- c. 7 X normal. (0.25) ]
- d. (five' required, 0.25 each) (1.25)
Chimney isolation valve closes. (5406) Air ejector suction valve closes. Mechanical vacuum pump trips. Off Gas sample vacuum pump suction valve closes. Off Gas line drain valve closes. (5408) i REFERENCE Quad Cities Off Gas (OG) system, pages 6, 14, 22. Quad Cities Off Gas Learning Objectives #2, #5. 271000G007 271000K406 2'100K404 ..(KA's) ANSWER 3.21 (1.00) c. l' l l REFERENCE Quad Cities ACAD/ CAM system, page 14. Quad Cities ACAD/ CAM Learning Objectives, #7. 272000A101 ..(KA's) . (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) f-c _ - - - -_ -
- 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page103 RESPONSIBILITIES (10%)
ANSWER 3.22 (1.00) a. REFERENCE Quad Cities Feed and Condensate system, page 14. Quad Cities Feed and Condensate system Learning Objective #4. 256000A201 ..(KA's) 1 ANSWER 3.23 (1.50) I a. Scram inlet ( 0. 2 5) and outlet valve are open (0.25) (for control rod 34-27).
- b. Low nitrogen pressure (<950 psig) (0.25) or high accumulator water level (>37 cc past accumulator seal.) (0.25)
- c. Rod at odd-numbered reed switch (0.25) (or rod not at even-numbered latch position) and no motion requested (0.25).
REFERENCE Quad Cities RMCS & RPIS page 4. QOA 900-5-G. Quad Cities RMCS & RPIS Learning Objectives, #6. ; 214000A301 214000A402 ..(KA's) i l l ANSWER 3.24 (1.00) j 1 b. REFERENCE Quad Cities Reactor Vessel and Internals, page 14. Quad Cities Reactor Vessel and Internals Learning Objectives, #4, #9. 290002K403 . . ( st.A ' s ) (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
3 '3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page104. , RESPONSIBILITIES (10%) ! > ANSWER 3.'25 (1.00)
- a. 1345'psig (0.5)'
b.- vessel steam; dome (top of the reactor vessel) (0.5) -REFERENCE Quad Cities Technical Specifications Bases, 1.2 . Quad Cities Reactor Vessel.and Internals, page 2. Quad Cities Learning Objectives, (not covered). 290002K507 ..(KA's) ANSWER 3.26 (1.00) I
- d. ,
REFERENCE Quad Cities, Tagging Equipment, QAP 300-13, page 4, 294001K102 ..(KA's) l ANSWER 3.27 (1.00) a. REFERENCE Quad Cities Equipment Out-of-Service, QAP 300-14, page 6. 294001K102 ..(KA's) (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
4
- 3. PLANT SYSTEMS 1(38%) AND PLANT-WIDE GENERIC Page105-RESPONSIBILITIES (10%) l l ANSWER. 3.28 -(1. 5 0 )
c, s' i a. Quarterly dose.~(1250 mrem - 1000. mrem). (3-75)
- b. MaximumLpermissible dose / dose rate = Stay time- l0.5) l i
250 mrem /40 mrem /hr= 6.25' hours (6 hours 15 minutes) (0,5) 1 NOTE: Any value of Maximum permissible dose:in part "b" is acceptable if-l -it is applied' correctly and is consistent with the final numerical I value. 1
, REFERENCE j Quad Cities' Radiation Protection Standards,.QAP 1000-1.
'294001K103 ..(KA's) i ANSWER 3.29 (1.00) 1
- d. i REFERENCE QAP 300-13 PG 1 AND 2. )
QAP 300-14 PG 1. .l 294001K102(3.9/4.5) ...K/A (K/A VALUE) l 294001K102 ..(KA's) i j
~l 1* )
! . ANSWER. 3.30 (1.00) ) b. I i , i [ REFERENCE L - L Quad Cities ACAD/ CAM, page 2. l 294001K115 ..(KA's) - y (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) 11 g
- - - _ _ _ _ _ _ _ _ . . - . a
3, PLANT SYSTEMS (38%{ AND PLANT-WIDE GENERIC Page106 RESPONSIBILITIES (10%) i I.NSWER 3.31 (2.00) ) l (0.5 each) (2.0) Sound the fire siren. I Announce over the page the location of the fire. Resound the fire siren. Maintain pressure on the fire protection water system. I l REFERENCE ) Quad Cities, Fires QAP 1170-3, page 1. l 294001K116 ..(KA's) ] l i j ANSWER 3.32 (1.00) 1 c. RE"ERENCE Quad Cities Operating Valve and Breaker Checklist, QAP 300-18, page 2. 294001K101 ..(KA's) ANSWER 1.33 (1.00) 1
- b. l
. REFERENCE l 10 CFR 20.
L 294001K103 ..(KA's) (***** END OF CATEGORY 3 *****) - (********** END OF EXAMINATION **********) l l L________-_____
I U. S. NUCLEAR REGULATORY COMMISSION l SENIOR REACTOR OPERATOR-LICENSE EXAMINATION REGION 3 ! FACILITY: Quad-cities 1 & 2 BWR-GE3 REACTOR TYPE:-
. DATE ADMINISTERED: 89/02/06 l
INSTRUCTIONS TO CANDIDATE: Use. separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. -Tne passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S' CATEGORY l VALUE TOTAL SCORE VALUE CATEGORY l
24.00 24.00 4. REACTOR PRINCIPLES (7%) . THERMODYNAMICS (7%) AND COMPONENTS (10%) (FUNDAMENTALS ) EXAM) l 33.00 33.00 5. EMERGENCY AND ABNORMAL PLANT l EVOLUTIONS (33%) 43.00 43.00 6. PLANT SYSTEMS (30%) AND l PLANT-WIDE GENERIC { RESPONSIBILITIES (13%) 100.0 % TOTALS FINAL GRADE All work done on this examination is my own. I have neither given nor received aid. Candidate's Signature a COPY L > MASTE1
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: ' 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination j room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil only to facilitate legible reproductions. j
- 4. Print your name in the blank provided on the cover sheet of the I examination. l t
- 5. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers. d l
- 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer sheet.
- 9. Number each answer as to category and number, for example, 1.4, 6.3. ;
- 10. Skip at least three lines between each answer. l
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAFK.
- 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
1
- 18. When you complete your examination, you shall:
- a. Assemble your examination as follows:
(1) Exam questions on top. (2) Exam aids - figures, tables, etc. (3) Answer pages including figures which are part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of the paper that you did not use for annwering the questions.
- d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
J l l
-- 2.-- m____m__-_. _ - _ _ _ _ _ _ _ _ _ _ _ -
l
- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 5 (33%)
l l I (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
m
- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 6' l (33%)' i
-l l
1 QUESTION. 5.01 (1.75) . The reactor is operating at 20% rated thernal power when reactor water level begins rapidly increasing. Answer the following questions in accordance with QOA 201-8, High or High High Reactor Water: Level,
- a. Why is the operator directed.to maintain reactor water level-below
+60' inches? (0.50) )
- b. What are FIVE-(5) COMPONENTS (SYSTEMS) that may suffer major damage due to excessively high vessel level during power operation? -(1.25) 4 I
l I J I J I l i (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) l-1 f.-
- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 7 (33%)
i
-QUESTION 5.02 (1.75) i l The reactor'is operating at 85% of rated power with the Recirculation 1 System in Master Manual Control when Recirculation Pump A speed rapidly !
increases to 104%, but a reactor scram does NOT occur. j Answer the following questions concerning QOA 202-2, Reactor Recirculation l System Failure - Flow Controller Fails High. l a. What is the Technical Specification BASIS for the immediate operator action requiring reactor recirculation pump-speeds to be within 10% of l each other? (0.75) I
- b. What are TWO'(2) IMMEDIATE ACTIONS that the operator is directed to' perform if he is unable to adjust recirculation pump speeds to within 10% of each other? (1.00) j q
, i l l l l i f (***** CATEGORY 5 CON 7INUED ON NEXT PAGE *****) l L . - _
'5 . EMERGENCY AND ABNORMAL PLANT EVOLUTIONS 'Page -8 (33%)
QUESTION 5.03 (2.00) -] The reactor,is operating at 90% rated: power when a loss of feedwater ! heating results in a 50 deg. F decrease in the feedwater injection temperature. Answer.the following questions concerning QOA 400-1, . Reactivity Addition, and QOA 3500-1, Loss of Feedwater Heaters,-'immediate operator actions,
- a. What is the BASIS for the immediate operator action which directs the operator to reduce power by reducing recirculation flow PRIOR TO x.
inserting control rods? (0150) l
- b. How'is the operator directed to INSERT control rods?
(FOUR (4) STEPS REQUIRED. )- (INCLUDE in your discussion ANY Protective i Features which must be overcome and ANY verifications which must be I made) (1.00) l
- c. When should~ control rod insertion be stopped? (0.50) l l
l l l I l 1 1 ! i (***** CATEGORY 5 CONTINUED ON NEXT PAGE'*****) l i
-1
- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 9 (33%)
~!
QUESTION 5.04 A small LOCA has caused a reactor scram. Reactor water level is steady at 50 inches and reactor pressure is 800 psig, decreasing slowly.. (5 psig/ minute), and Core Spray has automatically initiated due to a high drywell pressure. How will the Core Spray loop B subsystem respond if the associated minimum flow valve (MO-1402-38B) fails CLOSED? (INCLUDE a description of'the Core i Spray loop B flowpath in your discussion.) (1.00)
.1
'kt
-]
l (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) 9
- 5. EMERGENCY AND' ABNORMAL PLANT EVOLUTIONS Page 10 (33%)
1 QUESTION 3.05 (1.50)
.i The reactor is operating-at 100% rated power when-an off-gas explosion occurs causing the radioactivity release rate of Technical Specification j 3.8.A-to be exceeded.
Answer'the following. questions concerning QOA 5400-2, Off-Gas Explosion -
.Recombiner Operating and Filters Bypassed.
a.- What is the REASON for the'immediate action which directs the operator to SLOWLY reduce load at 20 MWe/hr in order to reduce the radioactive release below the Technical Specification limit? (TWO (2) REASONS REQUIRED.) (1.00)
- b. The immediate actione direct operators to verify the Off-Gas system integrity following the explosion. What RADIOLOGICAL CONTROLS should operators exercise when approaching the steam jet-air ejector rooms?
(0.50) { l i i
. . i
(***** CATEGORY 5 CONTINUED ON NEXT PAGE.*****) i ! _q l
- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 11 (33%)
QUESTION 5.06 (2.00) The unit 2 reactor is operating at 100% load with the recirculation system flow control in Master Manual when main condenser vacuum begins decreasing at 0.5 inches Hg vacuum per minute. What are EIGHT (8) of the nine IMMEDIATE OPERATOR ACTIONS (or automatic actions) that the Shift Engineer must verify? (2.00) i l l 1 1 (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) E____________
- 5. EMERGENCY AND ABNORMAL' PLANT EVOLUTIONS Page 12 (33%)
QUESTIOd 5,07 (3.00) An:wer the following questions concerning the Emergency Operating Procedures.
- c. What is the BASIS for General Caution number 3, which directs the operator to determine Suppression Pool temperature by averaging the channel A and Channel B temperatures of recorder 1640-9 on panel 901-21. (0.75)
- b. General Caution number 4 provides the following information for determining whether the reactor water level instruments may be used for vessel level indication.
Drywell T4.mperature Indication Level Reactor water Level Instrument l 361 deg F -309 inches GEMAC Lower 400 Range l (-334 to +66 inches) What CONDITION (S) must be met (using the Table above) to prevent use of the GEMAC Lower 400 Range water level instrument for level I indication? (0.50)
- c. General Caution number 9 provides the following conversion chart for
_ determining actual vessel level using the GEMAC Lower 400 Range. Reactor Pressure Actual Vessel Indicated Level (Inches) Level (inches) 500 psig Rx. Press 1000 psig Rx. Press
+30 -41 -66
+10 -57 -80
-60 -113 -130
-140 (~ Top of Core) -177 -188
-240 -257 -260
- 1. Why is the GEMAC lower 400 Range indicated level less than the actual vessel water level at reactor pressures of 500 psig and 1000 psig? (0.75)
- 2. What are the TWO (2) CONDITIONS which must be met in order to use thiJ conversion chart? (1.00)
(***** CATEGGRY 5 CONTINUED ON NEXT PAGE *****)
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- L _- -
- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 13 (33%)
QUESTION 5.08 (3.25) During a Unit 2 reactor plant transient, the following conditions exist: Suppression Pool level is 11.5 feet Suppression Pool average temperature is 145 deg. F , Drywell pressure is 10 psig Reactor pressure is 800 psig Suppression Chamber pressure is 9.75 psig Answer the following questions concerning the Heat Capacity Temperature Limit (HCTL) and the Heat Capacity Level-Limit (HCLL) curves.
** FIGURE QGA 100-F1/QGA 200-F1, HCTL, AND ** .
** FIGURE QGA 200-F7, HCLL, ARE ATTACHED. **
l and S tate
- a. DetermineAwhether the plant IS cr IS NOT within the limits of the Heat Capacity Level Limit for the plant conditions given above. (PLOT the point at which the plant is operating on the attached HCLL curve AND SHOW ALL WORK.) (1.25)
- b. What is the BASIS of the Heat Capacity Temperature' Limit (QGA FIGURE 100-1/QGA 200-F1)? (1.00)
- c. What is the BASIS for the MINIMUM allowed level of 10.9 feet for the Heat Capacity Level Limit curve (FIGURE QGA 200-F7) AND what ADVERSE CONDITION could occur if Suppression Pool level is less than 10.9 feet.
l (1.00) l 1 (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
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.5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page'14 (33%)
. QUESTION 5.09 (3.00)
Answer the following questions concerning QGA 100-3, RPV Power Control, a.- When is the operator required to enter QGA 100-3, RPV Power Control? (0.50)
- b. What is the BASIS for step B-3,'which directs the operator to " Confirm or initiate a recirc pump runback to minimum speed" prior to tripping-both recirculation pumps in step C-3? (INCLUDE ADVERSE CONSEQUENCE OF.
TRIPPING RECIRCULATION PUMPS WITHOUT REDUCTION IN SPEED.) (1.00)
- c. Why does step M-1 direct the operator to "Close 301-25 (CRD charging isolation valve". , (0.50)
- d. Why does step H-5 direct the. operator to " Inhibit ADS by placing the auto blowdown inhibit switch'to INHIBIT"? (0.50)
- e. Why,do steps F-5 and.H-5 direct the' operator to " Inject SBLC" if the l_ reactor CANNOT be shutdown before Suppression Pool temperature exceeds l 110 deg. F? (0.50) 1 i
J (***** CATEGORY S CONTINUED ON NEXT PAGE *****) '! - . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ . _ . _ _._ ._ ._ . _ _ _ _ ______.__.__.____._________._____.___________G
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- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 15 '
(33%)
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l QUESTION 5.10 (1.00) MULTIPLE CHOICE (Select the correct answer.) Select the ONE statement which explains why step D-3 of QGA 200-3, Primary Containment Pressure Control, PREVENTS the operator from j initiating Suppression Chamber sprays when Suppression Chamber pressure l is ABOVE 12.5 psig,
- a. To prevent excessive thermal shock to the Suppression Chamber spray I ring. 1 l
- b. To prevent depressurizing the Suppression Chamber so rapidly that the negative design pressure of the Suppression Chamber is {
exceeded.
- c. To prevent raising the level of the Suppression Chamber above the Drywell to Suppression Chamber vacuum breakers.
I d. To prevent chugging at the downcomer discharge. 1 I l l , l l l (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 16 (33%)
QUESTION 5.11 (1.00) I MULTIPLE CHOICE (Select the correct answer.) 'I
-Select the ONE statement which explains why step E-3 of QGA 200-3, Primary )
Containment ' Control, limits the operator to using only ONE (1) pump'for 1 Drywell Sprays . )
- a. To minimize damage to the equipment in the drywell due to the force l of the. water spray impingement.
l i b. To prevent flooding of the drywell at a rate that exceeds the capacity of the drywell sump pumps. To limit the speed with which the containment atmosphere is c. condensed. 1
- d. To prevent removal of all noncondensibles from the drywell atmosphere.
1
,a_.-
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 17 (33%)
l QUESTION 5.12 (1.50) Answer the following questions concerning step I-4 of QGA 200-3, Primary Containment Control. 1
- a. Why is the operator directed to vent the Primary Containment by taking I
a suction on the Suppression Chamber (with Standby Gas Treatment) when Suppression Pool level is BELOW 30 feet? (0.75)
- b. Why is the operator directed to vent the Primary Containment by taking '
a suction on the drywell (with Standby Gas Treatment) when Suppression Pool level is ABOVE 30 feet? (0.75) l i l l l (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
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- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 18 (33%)
i ( QUESTION 5.13- (1.00) MULTIPLE CHOICE (Select the. correct answer.) QGA 500-6, RPV Flooding, step L-4 states that the Minimum Flooding Interval with four (4) ADS valves open is 70 minutes (with RPV pressure,at least 77 ' psig above Suppression Chamber pressure.) Select the ONE statement which correctly completes the following statement. l The Minimum Flooding Interval is the amount of time necessary ... l , I
- a. to fill the RPV from the Bottom Reactor Head up to the ADS valves,
- b. to fill the RPV from the Bottom Reactor Head-up to the Top of Active Fuel.
- c. for a sufficient amount of decay heat to be removed such that injection systems may be secured for. lowering reactor water level in an attempt to restore level indication,
- d. to reduce fuel cladding temperature to less than 2200 deg. F )
following complete uncovery of the core. j I l l l l l (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
- 5. EMERGENCY AND ABNORMAL PLANT' EVOLUTIONS Page 19 (33%)
QUESTION 5.14 (2.00) Unit 2 reactor plant is operating at 75% rated power when a loss of the number 2 125V DC power supply occurs. (i.e. number 2 125V Battery is lost.) Answer the following in accordance with QOA 6900-4, Loss of Unit 2 125V DC Supply,
- a. MULTIPLE CHOICE (Select the correct answer.)
Select the ONE SIGNAL (METHOD) which will trip the Unit 2 Main Turbine. (1.00)
- 1. Depressing the turbine trip pushbutton in the control room
- 2. A high vibration trip signal
- 3. Operating the local turbine trip lever at the turbine. front standard
- 4. A Main Generator trip
- b. MULTIPLE CHOICE (Select the correct answer.)
Which ONE of the following Unit 2 systems will NOT automatically function during the loss.of 125V DC power. (1.00)
- 1. ADS (Auto Depressurization System)
- 2. Unit 2 Emergency Diesel Generator
- 3. HPCI (High Pressure Coolant Injection)
- 4. RCIC (Reactor Core Isolation Cooling)
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS ~Page 20 (33%)
i I, QUESTION 5.15 (1.00)
)
MULTIPLE CHOICE (Select the correct answer.) A loss of Shutdown Cooling has occurred on Unit 1. The reactor has.been ! shutdown for the past 12 hours, both recirculation pumps have been shutdown, the head vents have been opened, and shutdown cooling had been in i service for one hour. Which ONE of the following plant conditions would be indicated when the operator observes the upper vessel area metal temperatures increasing. (ASSUME NO OPERATOR ACTION IS TAKEN.)
- a. Proper natural circulation is occurring
- b. The. core is being cooled by reverse flow through the core
- c. Adequate core cooling is assured by steam cooling through the l head vents j
- d. Temperature stratification of the vessel coolant is occurring 1
l (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) __L_________ _-
-l S. - EMERGENCY AND ABNORMAL PLRNT EVOLUTIONS Page 21 (33%)_
QUESTION 5.16 (2.50) A Unit i reactor startup is in progress and reactor power is 10%'of rated when a leak in the Unit 1 Instrument Air header causes the Instrument Air pressure to STEADILY DECREASE.
- a. SELECT the ONE correct answer for each of the blanks'((1) through (3))
from the corresponding column listed below for the following statement. The low instrument air LOW pressure alarm will be received when air i pressure has decreased to (1) psig, the service air backup stop ! ("Little Joe") valve will open when instrument air pressure decreases j to (2) psig and the instrument air dryer bypass valve will open I at (3) .psig. (0.50 each) (1.50) j COLUMN (1) (2) (3) (a) 100 (a) 105 (a) 110 (b) 87 (b) 95 (b) 100 , (c) 77 (c) 85 (c) 90 I (d) 70 (d) 75 (d) 80 MULTIPLE CHOICE (Select the correct answer)-
- b. Which ONE of the following valves will fail CLOSED 'on a COMPLETE LOSS of Instrument Air. (1.00)
- 1. CRD flow control valve I
- 2. Feedwater regulating Low flow control valve
- 3. Reactor Building to Torus Vacuum Breakers (1601-20A and -20B) l 1
- 4. Reactor Feed Pump recirculation (minimum flow) valve 1
l l l l l (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
S. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 2'2 (33%)-
. QUESTION 5.17 (1.75)
Unit i reactor-is operating at 20% rated power when the following equipment j malfunctions are discovered: l - RCIC system is tagged out to repair a severe turbine governor valve steam leak. i - Electromatic Relief Valve 203-3B functional test shows that it will not receive an open signal when conditions are met for-ADS initiation RHR Service Water pump B motor tripped on ground fault and-cannot be started-
- All other Unit 1 equipment is fully operational 00 SELECTED TECHNICAL SPECIFICATIONS ARE ATTACHED **
j a. What is the most restrictive Limiting Condition for Operation ACTION STATEMENT which must be satisfied? (BE SPECIFIC AND STATE ANSWER IN. WORDS.) (0.75)
- b. State the Technical Specification surveillance (testing) requirements which must be satisfied. (BE SPECIFIC.) (1.00) i 1
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) ;
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- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 23 )
(33%1 { 1 QUESTION 5.18 (2.00) i MATCHING Match the Limiting Safety System Setting (LSSS) (in the Left-Hand Column) with the correct Basis (in the Right-Hand Column). (EACH ITEM IN THE LSSS (LEFT-HAND COLUMN) HAS ONLY ONE ANSWER AND EACH ITEM IN THE BASES COLUMN (RIGHT-HAND COLUMN) MAY BE USED MORE THAN ONCE OR NOT AT ALL.) (2.00) LSSS BASES I
- a. MSIV closure on 1. Prevents release of fission a l
low reactor pressure products due to fuel pin failure l l b. Condenser Low Vacuum 2. Prevents exceeding the MCPR I Scram Fuel Cladding Safety Limit {
- c. APRM 15% power scram 3. Prevents exceeding the LHGR Fuel Cladding Thermal Limit
- d. MSIV maximum closure time (5 seconds) 4. Anticipates main turbine stop valve closure l 5. Prevents exceeding 2200 deg. F peak fuel centerline temperature.
l l l (***** END OF CATEGORY 5 *****) I _ - _ - - - - - - - - - - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -J
- 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC Page 24 I
RESPONSIBILITIES (13%)
.i l
I J l I i l l 1 l I (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
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- 6. PLANT-SYSTEMS'(30%) AND PLANT-WIDE GENERIC Page 25 RESPONSIBILITIES (13%)
QUESTION 6.01 (2.75) As a result of an Anticipated Transient Without Scram (ATWS), it has.become necessary to initiate Standby Liquid Control (SBLC).- D. What are FIVE (5) of the seven positive reactivity contributions. which must be overcome by the SBLC system? (0.25 each) (1.25)
- b. Given the positions of SBLC system initiation switch as being
" system 2&l", " system 2", "off", " system-1", and " system'1&2", how does the operator position this switch to initiate SBLC. ( 0. 50 )'
- c. What.are.four (4) of the seven (7) parameters checked ta) verify.SBLC-initiation? (0.25 each) (1.00) 1
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
- 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC Page 26 RESPONSIBILITIES (13%)
QUESTION 6.02 (1.00) l MULTIPLE CHOICE (Select the correct answer.) A reactor startup is in progress and power indicates 10^3 cps on the source Range Monitors (SRM's). A malfunction occurs such that +\- 24 VDC power from the batteries to SRM A is interrupted. What is the impact of this malfunction?
\
l a. SRM A Drive mechanism Motor Module inoperative. l l l
- b. Ionization of Argon gas in SRM A ceases. l 1
- c. SRM A short period" alarm.
I d. SRM A inoperable. l 1 I
)
l 9 l l l l l l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
- 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC Page'27 RESPONSIBILITIES (13%)
QUESTION 6.03 (1.00) MULTIPLE CHOICE (Select the correct answer.) SRM's provide rod blocks to aid'in preventing the operator from adding ex- ) cessive reactivity to the core during a startup. Which'ONE (1) of the conditions.below will cause all SRM rod blocks to be i bypassed? l
- a. All IRM's on range 3 or above.
- b. SRM's read greater than 100 counts.
- c. All IRM's are on range 8 or above.
- d. SRM's read greater than 2 X 10^5 counts.
i l l l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) j
i l
- 6. ' PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC Page 28 RESPONSIBILITIES (13%)
1 QUESTION 6.04 (2.75) 1 The High Pressure Coolant Injection (HPCI) system automatically initiates i on low-low reactor water level (-59")-following a small instrument line-break. Reactor water level reaches +48" causing a HPCI turbine. trip just as a high HPCI area temperature alarm (200 degrees F) causes a system isolation,
- a. What is the other HPCI automatic initiation signal? (INCLUDE SETPOINT)
(0.25 initiation, 0.25 setpoint)- (0.50)
- b. -What are the remaining two (2) signals which will cause a Group IV system isolation? (INCLUDE UETPOINTS) (0.25 for signal, 0.25 for l setpoint). (1.00) !
- c. What are the three (3) DIRECT automatic actions which occur on a HPCI system isolation? (0.25 each) (0.75) 1
- d. How is the HPCI system isolation reset? (0.50) 1 i
l l l l l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
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- 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC- . Page 29 RESPONSIBILITIES (13%) I 4
QUESTION 6.05 (1.50) f l
.The reactor is operating at 100% rated power. Two (2) narrow range pressureLcompensated GE\MAC's indicate reactor water level from 0" to 60".
A malfunction occurs such that the GE\MAC's are no longer pressure compensated. Reactor pressure increases and ACTUAL water level remains constant. l
- a. What happens to indicated GE\MAC water level? -(0.50)
- b. . Why are all reactor water level trip signals obtained?from the Yarway instead of GE/MAC instruments? (0.50) .
- c. What is the purpose for the auxiliary chamber on the Yarway instruments? (see figure 3.1) (0.50) l l
i I l l j l l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) i i
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- 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC Page 30 RESPONSIBILITIES (13%)
l i 1 QUESTION 6.06 (1.00) MULTIPLE CHOICE (Select the correct answer.) i A small LOCA has occurred, causing the Emergency Diesel Generators (EDG) to ! auto start on high drywell pressure. EDG #1 trips shortly after startup. Which of the following would cause EDG #1 to trip?
- a. High jacket water temperature,
- b. Engine overspeed. ]
- c. Low lube oil temperature.
- d. Generator overvoltage.
l l l l l l l l l l i l I l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
.___.__._________________________U
- 6. ' PLANT SYSTEMS'(30%) AND PLANT-WIDE GENERIC Page 31 RESPONSIBILITIES (13%)
QUESTI:ON 6.07 (1.00) MULTIPLE CHOICE (Select the correct' answer.)' A reactor startup is in progress and power'is less than 10%. The Nuclear Station Operator (NSO) is withdrawing a control rod when a " SELECT BLOCK" is received. What is a possible cause for this alarm? 1
- a. CRD withdrawn past position "48"
- b. High water level in Scram Discharge Volume.
- c. Timer malfunction.
- d. APRM downscale (3/125 full scale).
l I i l l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) l l l . . i
- 6. PLAMT SYSTEMS (30%) AND PLANT-WIDE GENERIC Page 32 RESPONSIBILITIES (13%)
QUESTION 6.08 (2.75) The reactor operator intends to increase recirculation' pump speed from L twenty (20) percent to fifty.(50) percent; however the speed limiter in the E Recirculation Flow Control system will limit pump motor speed to 30% unless the pump discharge valve is open and feedwater flow is greater than twenty percent (20%) of rated.
- a. What are the power supplies to the reactor recirculation pump drive
! motors? (0.25 each) (0.50) 1
- b. What is the reason for limiting pump speed with the discharge valve shut? (0.50) l
- c. What is the reason for limiting pump speed with with less than 20% 1 feed flow? (0.50)
- d. How does the reactor recirculation system respond to an Anticipated Transient Without Scram (INCLUDE'ALL applicable trip signals,.
I setpoints, and recirculation system responses; 0.25 each)? (1.25) ) l 1 i i l l Il i (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
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- 6. PLANT SYSTEMS (30%) AMD PLANT-WIDE GENERIC Page 33 l RESPONSIBILITIES (13%)
QUESTION 6.09 (2.75) The Reactor Water Cleanup System is in the Blowdown Mode to control reactor water level during a reactor startup when an Anticipated Transient Without Scram (ATWS) occurs. When the operator initiates Standby Liquid Control, one (1) of the four (4) ways in which the RWCU system responds is a closure of the inboard suction / isolation valve.
- a. What are three (3) additional responses of the RWCU system?
(0.25 each) (0.75)
- b. What two (2) other conditions will cause the RWCU system to isolate?
(INCLUDE ISOLATION SIGNAL AND SETPOINT) (0.50 each) (1.00) i
- c. How will the Blowdown Flow Control Valve respond (open fully, close fully, remain as is, throttle open) following the initiation of SBLC?
EXPLAIN YOUR ANSWER. (1.00) l l 1
)
l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
-__ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ =
- 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC Page 34 RESPONSIBILITIES (13%)
QUESTION 6.10 (1.00) MULTIPLE CHOICE (Select the correct answer). The reactor is operating at 80% power with the following feed and condensate pump alignment: Feed pumps A and B running. Condensate Booster Pumps A, B, and C running with condensate booster pump D in standby. Condensate pumps A, B, and C running. When will the standby condensate pump start?
- a. When feed pump suction pressure is less than 160 psig.
- b. When condensate flow decreases below 95% full flow.
- c. When the A condensate pump breaker opens.
- d. When hotwell level increases to 33 inches.
l (***** CATEGORY G CONTINUED ON NEXT PAGE *****)
. i.
- 6. PLANT SYSTEMS (30%) AND PLANT-WIDE-GENERIC Page 35 RESPONSIBILITIES (13%)
QUESTION 6.11 (1.50) The operator is observing the Full Core Display while withdrawing control rods for a reactor startup. What is indicated ^by each of the following lights on the. Full Core Display? (Two conditions required for each, 0.25 each.)
- a. A' BLUE LIGHT for rod 34-27 (0.50) l
- b. An accumulator trouble light for rod 16-25. (0.50)
- c. A red rod drift light for rod 18-23 (0.50) 1
! (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
- _ _ - - =. - _ _ _ _
- 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC Page 36 RESPONSIBILITIES (13%)
QUESTION 6.12 (1.00) MULTIPLE CHOICE (Select the correct answer) . The amount of two (2) phase flow resistance within a fuel bundle increases as power increases. This hao a tendency to re-route flow to the lower power bundles. Specific design features have been incorporated to minimize the impact of increased flow resistance due to two (2) phase flow. Which selection represents one (1) of these design features?
- a. Placing lower enriched fuel at the periphery of the core.
j b. Installing flow orifices in the fuel support pieces. l l c. Minimizing the length of the active fuel region.
- d. Limiting Linear Heat Generation Rate (LHGR) to 13.4 KW/FT.
( l l 1 l l l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
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- 6. " PLANT SYSTENS (30%)'AND PLANT-WIDE GENERIC Page 37 RESPONSIBILITIES (13%)
1 l f QUESTION 6.13 (3.00) ! a) How do the Backup Scram Valves respond to a trip on RPS Channel A?. (Both valves-open, Backup scram valve A opens, Backup scram valve B opens, Neither Backup Scram valve opens.) . (0.50)
.b) What is the power supply to the Backup Scram Valve solenoids? (0.50)
I l c) What is the response of the following val'res if Backup Scram i Valve B is inadvertently opened during testing? (OPEN,-CLOSE,- l REMAIN OPEN, OR REMAIN CLOSED) i
- 1. Scram pilot solenoid valves. (0.50)-
l 1 2. Inboard Scram Discharge Volume drain valves (0.50)
- 3. Scram inlet valves - ( 0 .' 5 0 )
- 4. ATWS scram solenoid valves (0.50) l l
i l l 1
)
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- 6. PLANT SYSTEMS (30%) AMD PLANT-WIDE GENERIC Page 38 RESPONSIBILITIES (13%) ;
i QUESTION 6.14 (2.00) A reactor shutdown is in progress and the operator is inserting control rods. IRM 12 is reading 11 on range 7 (0-40 scale). a) State the response of the plant, when the operator places the IRM 12 range switch to range 6. (rod block, half scram, full scram, or no action.) EXPLAIN THE PLANT'S RESPONSE. (1.00) b) As reactor power is reduced, IRM 15 indication decreases to a reading of 3 on range 1. What is the response of the plant? (rod block, half scram, full scram, or no action.) EXPLAIN THE PLANT'S RESPONSE. (1.00) 1 l l l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
'6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC Page 39 RESPONSIBILITIES (13%)
.l
.{
QUESTION 6.15 (2.50) l A surveillance is in progress on the Standby Gas Treatment (SBGT) system. SBGT fan A is running (selected to primary) and taking a suction on Unit 2
];
Reactor Building:(through MO-7503A) when a Unit 1 high Drywell pressure
' condition provides an automatic ~ initiation signal to the SBGT system.
'. 1 l a. Describe the response of the SBGT suction dampers to Unit 1 (MO-7503B) I and Unit 2 (MO-7503A) Reactor Buildings.. .(0.50)
- b. State FOUR (4) additional AUTOMATIC INITIATION SIGNALS for the !
'SBGT system. (INCLUDE SETPOINTS.) (2.00) i l
l
)
l l
)
1. (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) i ______ _ _ _ _ _ _ _ . _ . _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ . _ . - _ . _ _ _ m__ _ . _ _ - _ _ _ . _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _
_y _ _ - - - - - - _ - _ - - - - - - - - _ _ _ _ - - - _ _ _ - - _
- 6. PLANT SYSTEMS (30%)~AND PLANT-WIDE GENERIC Page 40 RESPONSIBILITIES (13%)
QUESTION 6.16 (2.50) l During a reactor transient.the operator scrams the reactor by placing the mode switch to SHUTDOWN. a) State TWO (2) of the three design purposes for a reactor scram. (1.00) l b) Why is the SCRAM SIGNAL generated by placing the mode switch'in the SHUTDOWN position automatically BYPASSED after 2 seconds? l (0.75) l c) Why is the operator prevented from resetting the reactor scram signal for 10 seconds after the mode switch is placed into SHUTDOWN? (0.75) l l i a l l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
- 6. PLANT SYSTEMS (30%)~AND PLANT-WIDE' GENERIC Page 41 RESPONSIBILITIES (13%)
i QUESTION 6.17 (1.00) MULTIPLE CHOICE (Select the correct answer.) i Under certain circumstances, it is permissible to perform only one . (1)- ! Valve position verification of a safety.related system during an operating valve lineup check. i Which selection represents one (1).of these circumstances?
- a. When authorized by the Nuclear Station Operator. )
- b. When remote indication is available.
- c. In. cases which involve significant. radiation exposure.
- d. When the valve is listed on the Out-of-service sheet.
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
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- 6. ' PLANT SYSTEMS-(30%) AND PLANT-WIDE GENERIC- -
Page'42
-RESPONSIBILITIES ( 13 %')
QUESTION 6.18 (1.00) 1 MULTIPLE CHOICE (Select the correct answer.) i s Which'ONE of the following cards is used to HOLD equipment out-of-service j for the protection of personnel working on equipment. j
- a. Personnel Protection card l i
- b. Caution card j
- c. Hold card i
- d. 'Out-of-Service card.
l l
- l 1
1 ( (***** CATEGORY 6 CONTINUED ON NEXT PAGE.*****)
- 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC Page 43 RESPONSIBILITIES (13%)
QUESTION 6.19 (1.00) MULTIPLE CHOICE (Select the correct answer.) Which ONE of the following is the LAST card that is hung when removing a piece of equipment from service for maintenance.
- a. Master Personnel Protection card
- b. Master Out-of-Service card
- c. Special Order card
- d. Quality Assurance Hold card 1
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- 6. PLANT SYSTEMS (30%)-AND PLANT-WIDE GENERIC Page 44 RESPONSIBILITIES (13%)
QUESTION 6.20 (1.00) ; I MULTIPLE CHOICE'(Select the correct answer.) A Unit i reactor plant transient is in progress and the Acting Station Director has declared an Unusual Event. If the Acting Station Director upgrades the event classification to an Alert, what is the time. period . within which.the STATE AUTHORITIES must be NOTIFIED. l
- a. 24 hours
- b. 4 hours
- c. I hour
- d. 15 minutes .f l
l I l (***** CATEGORY- 6 CONTINUED ON NEXT PAGE *****) l a_________-_____-_
- 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC Page 45 RESPONSIBILITIES (13%)
)
QUESTION 6.21 (1.00) ) A' RUPTURE in the MAIN. STEAM piping of Unit 2 reactor has just occurred. Plant conditions are as follows: i
- The reactor failed to automatically scram, and the operator initiated a manual scram which inserted all control rods.
- MSIVs isolated almost instantly Drywell temperature is 284 deg. F Drywell pressure is 52 psig 1
Suppression pool temperature is 145 deg F l Reactor water level dropped to -80 inches in the first minute and due to the failure of various ECCS systems LEVEL is being maintained at -70 inches. High radiation alarms on the Main Chimney SPING Mid Range have been received and a release rate of 4.0E6 uci/see has been calculated. 1 I Determine the NARS FORM boxes which must be checked for this event in accordance with the Quad-Cities Emergency Plan procedures, QEPs. (BE SURE TO STATE THE EMERGENCY CLASSIFICATION LEVEL. ) (1.00)
** QEP 100-T2, PREDETERMINED PARS FOR CONTROL ROOM, AND **
** QEP 200-T1, QUAD-CITIES EMERGENCY ACTION LEVELS, ARE ATTACHED **
l I i (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
.m s
- 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC Page 46 RESPONSIBILITIES (13%)
QUESTION 6.22 (1.50) An operator returns from two days off (Thursday and Friday) and works tha following shift hours as a control room operator during a Unit 1 outage: Saturday - 6 am to 4 pm (0600 to 1600) Sunday - 6 am to 2 pm (0600 to 1400) Monday - 6 am to 2 pm (0600 to 1400) Tuesday - 6 am to 6 pm (0600 to 1800) a) What is the earliest hour at which the operator can return and reassume the shift on Wednesday as a unit operator? (0.50) b) What is the maximum number of hours that this operator can work between 12 am and 6 pm (0000 to 1800) on Wednesday? (0.50) l l c) The operator works the following hours on Wednesday and Thursday: Wednesday - 6 am to 2 pm (0600 to 1400) Thursday - 10 am to 6 pm (1000 to 1800) What is the maximum number of hours that the operator can work on Friday? (Between 12:00 am and 12:00 am OR 0000 to 2400) (0.50) l I i (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
- 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC Page 47 RESPONSIBILITIES (13%)
)
QUESTION 6.23 (2.00) Answer the following concerning temporary procedure changes.in accordance i with QAP 1100-7, Approval and Authorization of Permanent Procedures. '
- a. FILL IN THE BLANK (1.00)
Temporary changes to PERMANENT procedures which CHANGE the INTENT of the permanent procedure may be made with the concurrence.of (1) i and'one of the individuals identified in the second-column of QAP / 1100-T1 and it must be approved by (2) prior to implementation. f I
- b. MULTIPLE CHOICE (Select the correct answer.) (1.00)
Which ONE of the following does NOT constitute a change to the
" INTENT" of a procedure.
- 1. The temporary change to a permanent procedure is more conservative than the permanent procedure;with respect to Technical Specifications.
- 2. The temporary change alters the content of two principal steps.in the startup of a safety related system.
- 3. The temporary change alters;the order of the principal steps for l
the performance of a Technical Specification surveillance. 1 *
- 4. The procedure is a new procedure being implemented with the temporary procedure request sheet.
i l 1 (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) l
.y
- 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC Page 48 RESPONSIBILITIES (13%)
i
' QUESTION 6.24 (1.50)
Answer the following questions concerning access control to Locked High Radiation Areas in accordance with QAP 1120-6, Entering a Locked High Radiation Area Without a Timekeeper, and QAP 1900-9, Lock and Key Control,
- a. FILL IN THE BLANK (More than one word may be required to complete the blank) (0.50)
An individual entering a high radiation area shall notify the Communications Center or the Shift Engineer / Shift Foreman immediately before entering the high radiation area unless a has been designated.
- b. State the MAXIMUM number of R-keys that can be issued to a rounds operator. .(0.50)
- c. How is positive access control to a high radiation area maintained 1when the R-lock must remain unlocked in order to run a welding cable into the area. (0.50) 1 l
i i i (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
I
)
- 6. ' PLANT SYSTEMS ( 3 0_%_) AND PLANT-WIDE GENERIC Page 49 RESPONSIBILITIES (13%)
QUESTION 6.25 (1.00) I Prior.to clearing the Drywell for controlled access, an INITIAL ENTRY into ! the Drywell must be made in accordance with QAP'1150-3, Initial Entry to the Drywell or Suppression Chamber. ONE (1) of the reasons for the INITIAL ENTRY is to confirm the results of the. atmospheric samples drawn through the Drywell Oxygen Analyzer. j State the remaining TWO'(2) of the three REASONS for the Initial Drywell Entry. (1.00) I I l I l 1 i l l ! l l l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) 1
"6. PLANT SYSTEMS'(30%) AMD PLANT-WIDE GENERIC Page 50- !
RESPONSIBILITIES (13%) J QUESTION 6.26 (2.00) I A reactor transient is in progress which has resulted in the Notification of an Unusual Event. J What are FOUR (4) of the six CONDITIONS which require the Station
. Emergency Director to make an IMMEDIATE follow-up notification to the NRC Operations Center. (2.00) 1 I
l l i l I i I I (***** END OF CATEGORY 6 *****) . _ (********** END OF EXAMINATION **********) i __ _ . - _ - . ___-_-_____:_w
1 [-
<v-- y
.c :k 1
o.. [ l': di.'.'? . [I b I i 1 l ANSWER KEY Quad-Cities 1989/02/06 SRO Examination - l l l l r -~m..-
; .i s e r e 6
~
- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 52 (33%) )
ANSWER 5.01 (1.75)
- a. To maintain level within the limits of his indication on panel 901-5 (902-5). (0.50)
- b. 1. HPCI turbine
- 2. RCIC turbine c coc, cte. g u m h v p w .
- 3. Main turbine
- 4. Main steam line piping
- 5. Reactor vessel
- 6. SRVs l (5 required, 0.25 ea.)
i . REFERENCE j l- 1 l QOA 201-8, PG 1 - 3. f l 295008A101(3.7/3.7) 295008K101(3.0/3.2) 295008K205(3.8/3.9) ! 295008K206(3.4/3.6) 295008K208(3.4/3.5) 295008K211(3.1/3.3)
...K/A (K/A VALUE) l 295008K208 295008K206 295008K205 295008K101 295008A101 l ..(KA's) l l
l I l ANSWER 5.02 (1.75)
- a. To prevent the LPCI Loop Select logic from selecting the wrong loop for injection (in the event of a LOCA). (0.75)
- b. 1. Trip Recirc pump A (0.50)
- 2. Place the controller for recirc pump B in manual (0.25) and reduce load (pump speed) to prevent recirc pump B overload (724 AMP).
(0.25) "
.3 c - 41, cA4., cep ^ Ce cicA 2. c 2 - 4 REFERENCE QOA 202-2 PG 1 AND QOA 202-4 PG 1.
TECH SPEC BASES PG 3.6/4.6-24, 295001A101(3.5/3.6) 295001A1059(3.3/3.3) 295001G010(3.8/3.7) 295001K205(3.2/3.6) ...K/A (K/A VALUE) 295001K205 295001G010 295001A105 295007.A101 ..(KA's) 1 j (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
--_.______,x---..-_--_____-.-_-_-__ ~_--___-_-___.-__.___-_._-_____.n_ _ . - - - _ _ _ _ .
- 5. EMERGENCY - AND ABNORMAL PLANT EVOLUTIONS _ Page 53 (33%)
l
. ANSWER 5.03 (2.00) ,
a.
- z. c c. Y< " ' m<(e < c$i/ > ~ clw th h l! * '
$ i .uAPRM flow;3 bias line has been selected t'o allow adequate margin to S /*
avoid a scram as flow is reduced. (reduced from 0.65w.tog 0.58w) (0.50),,w,,- ; (). . .q c -I: sc w- a n s; t > ,;. u w ..,w., ;t. c r u ,w i p Q[ta . ovy )
- b. 1. Place RWM mode switch to bypass b
- 2. Insert control rods in reverse sequence
- 3. Continuously insert to full-in (any rod inserted)
- 4. Check the flow control line after each group of rods inserted.
(4 required, 0.25 ea.)
- c. When the reactor is operating below the 100% flow control line. (0.50)
REFERENCE l QOA 400-1 PG 1 AND 3. QOA 3500-1 PG 2. 1 295014A104(3.2/3.3) 295014A204(4.1/4.4) 295014G007(3.3/3.6) 295014K104(3.0/3.4) 295014K209(3.4/3.6) 295014K211(3.6/3.7)
...K/A (K/A VALUE) 295014K209 295014K104 295014G007 295014A204 295014A104
..(KA's)
ANSWER 5.04 (1 , / lq 4' The core spray loop B discharge relief valve (RV-1402-28B) will lift and divert water to the RBEDT from the torus (also accept appropriate i discussion of this transfer of water.) .(1.00)
/fY QOA 1400-1 PG 2.
{l'k['
- h.,]'~ , [q[ 7'[yL 1 M '
^^ 't
' ,/
g,,/ s , y., t./ .4 U 295036A203(3.4/3.8) 295036K201(3.1/3.2) ...K/A ,,' j 295036K201 295036A203 ..(KA's) (K/A
.W VALUE)[.
Tj /m , fi -7' f A I (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) 1
- _______ _________ ___ _ _ j
- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 54 (33%)
l ANSWER 5.05 (1.50)
- a. The slow power change will minimize the increased release of long-lived fission products. (0.50) and will also minimize the increase in the release rate of radioactive l iodine (from the chimney and the reactor building vent stack). (0.50) !
- b. Full protective clothing. (0.50) (and full face mask) ]
(W4 b w N ae g/rM ) REFERENCE l QOA 5400-2 PG 1, 2, AND 4. 295038A101(3.9/4.2) 295038A204(4.1/4.5) 29503 8G007 (3. 2/3. 5) I 295038K102(4.2/4.4) 295038K302 (3.9 /4.2) ...K/A (K/A VALUE) 295038K302 295038K102 295038G007 295038A204 295038A101 l
..(KA's)
I ANSWER 5.06 (2.00) t i
- 1. Reduce reactor power (with recirculation flow / control rods to maintain vacuum) (0.25)
- 2. Check and fill condenser loop seals (0.25) l
- 3. Check SJAE for failure (send operator to SJAE) (0.25) i
- 4. Verify condenser vacuum breaker is closed (0.125) and has a water seal (0.125)
- 5. Notify Load Dispatcher (0.25)
- 6. Verify reactor scrams at 21 inches HG (mode switch is in run) (0.25)
- 7. Verify main turbine trip at 20 inches Hg vacuum (0.25)
- 8. Verify all turbine bypass valves close when vacuum < 7 inches Hg vacuum (0.25)
- 9. Open steam seal bypass if necessary to maintain seal pressure. (0.25)
(accept for partial credit: Verify turbine steam seal pressure is normal (0.125)) l l (8 required) 1 (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS- Page 55 (33%)
REFERENCE QOA 3300-2 PG 1 AND 2. 295002A103(3.4/3.5) 295002A105(3.2/3.2) 295002 A106 ( 3. 0/ 3.1) 295002G010(3.8/3.7) ...K/A (K/A VALUE) 295002G010 295002A106 295002A105 295002A103 ..(KA's) ANSWER 5.07 (3.00) l l
- a. The suppression pool is a very large volume where wide variations in temperature may occur, resulting in local temperatures which may a not be representative of the entire volume. (0.75) !
(OR Actions within the emergency operating procedures (QGAs) are based i upon bulk (average) suppression pool temperature.) :
- b. Drywell temperature exceeds 361 deg. F (0.25) and the instrument reads below -309 inches. (0.25)
I
- c. 1. GEMAC lower 400 is calibrated under cold (0 psig, 70 deg. F)
{ conditions. (0.75) q l
- 2. Recirculation pumps off. (0.50) i Drywell temperature (approximately) normal (0.50)
REFERENCE LESSON PL7N ON EMERGENCY OPERATING PROCEDURES PG 12, 14, AND 15. LEARNING OBJECTIVE NO. 4. 295026A103(3.9/3.9) 295026A201(4.1/4.2) 295026G007(3.4/3.8) 295028A203(3.7/3.9) 295028G007(3.4/3.8) 295028K101(3.5/3.7) 295028K203(3.6/3.8) ...K/A (K/A VALUE) 295028G007 295028A203 295026G007 295026A201 295026A103
..(KA's)
{ i (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) l L
- 5. EMERGENCY AND ABNORMAL PLRNT EVOLUTIONS Page 56 (33%) 9 ANSWER 5.08 (3.25)
- a. Delta T = HCTL - Suppression Pool Temperature HCTL = 159 deg. F .
(0.50) Delta T = 159 - 145 = 14 deg. F (0.25) Plot point at 11.5 ft and 14 deg. F on FIG. QGA 200-F7 (0.25) Therefore plant IS operating WITHIN HCLL curve (0.25)
- b. The HCTL defines the plant conditions at which the heat capacity of the suppression pool is sufficient to accommodate a blowdown of the RPV (0.50) to the shutdown cooling interlock pressure (0.25) without unstable steam condensation. (0.25)
- c. The minimum level is the level at which the downcomers are uncovered.
(0.50) With level < 10.9 ft overpressurization of the. containment could occur in the event of a LOCA . (0.50) - l l REFERENCE LESSON PLAN ON EMERGENCY OPERATING PROCEDURES PG 28 AND 29. LEARNING OBJECTIVE NO. 7. 295026A202(3.8/3.9) 295026G007(3.4/3.8) 295026K206(3.5/3.7) 295030A201(4.1/4.2) 29503 0A202 ( 3. 9/ 3. 9 );. 295030A203(3.7/3.9) 295030G007(3.6/3.9) '35030K103(3.8/4.1) 7 295030K207(3.5/3.8)
...K/A (K/A VALUE':
295030A202 2DSJ30A201 295026K206 295026G007 295026A202
..(KA's)
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- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Pago 57 (33%)
ANSWER 5.09 (3.00)
- a. Anytime that QGA 100-1, RPV Level Control, is entered. (0.50)
(Also accept: listing ALL entry conditions for QGA 100-1.)
- b. Provides for more control of the RPV level transient (0.50). Without the speed reduction a turbine trip (+48 inches) may occur due to the level swell (0.50).
- c. To prevent diverting flow (pressure) from the drive water header to allow maximum CRD pump discharge for driving control rods manually. >
(0.50) (Also. accept: To allow FCV to open and supply flow for driving control rods manually.)
- d. To prevent a cold water transient due to automatic injection of systems during depressurization. (0.50)
- e. To assure a reactor shutdown will be achieved before the HCTL of the suppression pool is exceeded. (0.50)
REFERENCE LESSON PLAN ON EMERGENCY OPERATING PROCEDURES PG 42 - 45. LEARNING OBJECTIVE NO. 6. 295037 A206 ( 4. 0/4.1) 295037G007(3.7/3.9) 295037G011(4.4/4.7) ! 295037K204(4.4/4.5) 295037K205(4.0/4.1) 295037K301(4.1/4.2) l 295037K302(4.3/4.5) ...K/A (K/A VALUE) 295037K205 295037K204 295037G011 295037G007 295037A206
..(KA's)
ANSWER 5.10 (1.00) l d. n L l l i I (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) , I i i
- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS .Page 58-(33%)
REFERENCE LESSON PLAN ON EMERGENCY OPERATING PROCEDURES PG 53. LEARNING OBJECTIVE NO. 6. 295024G007(3.6/3.9) 295024K213(3.8/3.3) 295024K302(3.5/3.8) 0 ..K/A (K/A VALUE) 295024K302 295024K213 295024G007 ..(KA's) ANSWER 5.11 (1.00) c. REFERENCE LESSON PLAN ON EMERGENCY OPERATING PROCEDURES PG 53. LEARNING OBJECTIVE NO. 6. 295024G007(3.6/3.9)' 295024K102(4.1/4.2) 295024K211(4.2/4.2) 295024K308(3.7/4.1) ...K/A (K/A VALUE) 295024K308 295024K211 295024K102 295024G007 ..(KA's) ANSWER 5.12 (1.50)
- a. So that some decontamination of the drywell atmosphere is performed as it passes through the suppression pool water to be vented. (0.75)
- b. _The vent tap from the suppression chamber will be covered with water.
(0.75) I REFERENCE l LESSON PLAN ON EMERGENCY OPERATING PROCEDURES PG 55. LEARNING OBJECTIVE NO. 6. 295024A203(3.8/3.8) 295024G007(3.6/3.9) 295024K307(3.5/4.0)
. ..K/A (K/A VALUE) 295024K307 295024G007 295024A203 ..(KA's) l l
l 1 (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
EMERGENCY AND ABNCRMAL PLANT EVOLUTIONS Page-59 1 (33Q ANSWER 5.13 (1.00) b. 1 REFERENCE LESSON PLAN ON EMERGENCY OPERATING PROCEDURES PG 81. LEARNING OBJECTIVE NO. 6. QGA 500-6, FLOW CHART. 295031A201(4.6/4.8) 295031G007(3.7/4.0) 295031K216 (4.1/4.1) '
...K/A (K/A VALUE) 295031A201 ..(KA's)
~
295031K216 295031G007 ANSWER 5.14 (2.00)
- a. 3. (1.00)
- b. 4. (1.00)
REFERENCE LESSON PLAN ON LEARNING OBJECTIVE NO. i QGA 6900-4 PG 1, 2, AND 5. 295004A102(3.8/4.1) 295004G007(3.0/3.5) 295004K203(3.3/3.3) , ...K/A (K/A VALUE) l 295004K203 295004G007 295004A102 ..(KA's) )
)
ANSWER 5.15 (1.00) i
- d. !
l (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
- - _ _ _ _--_---__ j
- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 60 (33%) .
l REFERENCE LESSON PLAN ON LEARNING OBJECTIVE NO.
-QGA 1000-2 PG 4.
295021A205(3.4/3.5) 295021K102(3.3/3.4) 295021K103(3.9/3.9) 295021K104(3.6/3.7) ...K/A (K/A VALUE) , 295021K104 295021K103 295021K102 295021A205 ..(KA's) ) l l ANSWER 5.16 (2.50) i a. (1) (b) 87 (2) (c) 05 (3) (d) 80 (3 required, 0.50 ea.) I
- b. 1. (1.00)
REFERENCE i LESSON PLAN ON CRD HYDRAULICS PG 22 LESSON PLAN ON FEED AND CONDENSATE PG 26. LEARNING OBJECTIVE NO. 4. QOA 4700-2 PG 1. 295019A101(3.5/3.3) 295019K201(3.8/3.9) 295019K203(3.2/3.2) 295019K301(3.3/3.4) ...K/A (K/A VALUE) 295019K301 295019K203 295019K201 295019A101 ..(KA's) ANSWFR 5.17 (1.75)
- a. Reactor operation may continue for the succeeding 7 days (0.75)
(T.S. 3.5.D.2 and 3.5.E.2) c, 50 c . ?%
- b. HPCJ gust be run (Or29) immediately- (-C.125) and daily thereaf ter
( . . - (T.S. 4.5.D.4), (C#2.-and-RHR loop B operable compon,ents and RHR loop A components for - l containment cor' ooling (&lst'must be'demoqq'trated operable immediately ( and daily thereafter (4.'125) (T.S. 4.5.B.2)) dW ;b , WW -S d w - e _% p N N ' g W q. . s
. % y p_iAT2. .
i I (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) 1 I
- 5. EMERGENCY AND' ABNORMAL PLANT EVOLUTIONS. Page 61 j (33%) ]
REFERENCE
. LESSON PIAN ON AUTO PRESSURE RELIEF SYSTEM.
LEARNING OBJECTIVE NO. 7. 295007G003(3.4/4.2) 295019G003(2.9/3.6) ...K/A (K/A VALUE) 295019G003 295007G003 ..(KA's) ANSWER 5.18 (2.00)
- a. 2
- b. 4
- c. .2 j
- d. 1 j i
(4 required, 0.50 ea.) REFERENCE TECHNICAL SPECIFICATIONS PG 1.1/2.1-12, -15, AND -16. TECHNICAL SPECIFICATIONS PG 3.7/4.7-32. . 295005G004(2.7/3.7) 295014G004(3.3/4.2) 295017G004(2.6/3.8)
...K/A (K/A VALUE) l 295017G004 295014G004 295005G004 ..(KA's) l i
.l l
-I i
1 i (***** END OF CATEGORY 5 *****) l
.I
- 6. PLANT. SYSTEMS (30%) AND PLANT-WIDE GENERIC Page 62 RESPO 47TBILITIES (13%)
ANSWER 6.01 (2.75) l l a. Elimination of steam voids. Moderator temperature change (from hot to 125 degrees F.) Reduced doppler. Decreased control rod worth as the moderator cools. Xenon decay. Shutdown margin 3% dK Reactivity associated with control rods not inserted (5 required, 0.25 ea.)
- b. Place the switch to the system 1&2 position (through position system 1)
(also accept: to the system 2&1 position through system 2 position) (0.50)
- c. Flow indicating pilot light lit Decreasing SBLC tank level SBLC pump discharge pressure > reactor pressure.
Neutron flux level decreasing. Reactor Water Cleanup System isolates. Squib continuity light goes out. Squib valve circuit failure annunciator. (4 required, 0.25 ea.) REFERENCE QOP 1100-2 Quad Cities Standby Liquid Control, page 2, 3, 10. Technical Specification, 3.3 . 211000A305 211000A302 211000A301 211000A408 21100K503
..(KA's)
ANSWER 6.02 (1.00) d. (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
- 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC .Page 63 RESPONSIBILITIES (13%)
2 REFERENCE Quad Cities Source Range Monitors, figure 1. Quad Cities Source Range Monitors Learning Objectives, #4. 215004G007 215004K201 21500K602 ..(KA's) J ANSWER 6.03 (1.00) l c.
)
)
REFERENCE l l Quad Cities Source Range Monitors, pages 13, 14. 1 Quad Cities Source Range Monitors Learning Objectives, #5. l 215004K401 215004K102 ..(KA's) ] 1 1 ANSWER 6.04 (2.75)
- a. High drywell pressure 2.5 psig (+\ .25psig)
I (0.25 per setpoint, 0.25 per signal)
- b. Low reactor pressure 100 psig (+\- 10 psig)
High steam line flow 300% (+\- 30%) (0.25 per setpoint, 0.25 per signal) 4 1 4, -d -*"""2
- c. Suppression pool suction line isolates.
Auxiliary oil pump interlocked against autostart. M'*,~y* cuas ce4+ 3" - "~~' Steam Supply line isolates. c~o ~ e "*y " *- (3 required, 0.25 ea.)
- d. Depress the " Reset valves 4+5" pushbutton. (0.50)
REFERENCE Quad Cities High Pressure Core Injection, page 4, 68. Quad Cities High Pressure Core Injection Learning Objectives:#5,#6, #8. 206000K407 206000K404 206000K402 206000K116 ..(KA's) (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
.T
- 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC ,
Page 64 RESPONSIBILITIES (13%) ANSWER 6.05 (1.50)
- c. Indicated. level decreases (due to a lower water density seen by the dp cell which increases the d/p output between the variable leg and the reference leg.) (0.50)
- b. Yarways require no power supply for local operation. (0.50)
- c. Helps maintain the reference leg full during rapid depressurization of-the reactor vessel. (0.50) l REFERENCE 1
i Quad Cities Reactor Vessel Instrumentation, pages 2, 12, 14. Quad Cities Reactor Vessel Instrumentation Learning Objectives #1, #3. 216000K513 216000K324 216000K301 216000K112 ..(KA's) ANSWER 6.06 (1.00) b. REFERENCE l Quad Cities Diesel Generator, page 46. Quad Cities Diesel Generator Learning Objectives, #5. l [ 264000K402 ..(KA's) ANSWER 6.07 (1.00) c. REFERENCE
~
Quad' Cities RMCS & RPIS, page 20. Quad Cities RMCS & RPIS Learning Objective 4. 201002K301 ..(KA's) (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
--x__---. - - - - - . - - - . - - - Page 65 j
- 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC RESPONSIBILITIES (13%)
I i ANSWER 6.08 (2.75) I
- a. Bus 11 Bus 12 I
(0.25 each)
- b. Excessive axial thrust could be developed across the pump. (0.50)
- c. Prevents pump damage due to cavitation. (0.50)
- d. A high reactor pressure (0.25) of 1250 psig (0.25), energizes the tripp' coils for the recirculation pump motor generator field trip coils (0.25), low reactor water level (0.25) of -59 inches. (0.25)
REFERENCE Quad Cities Recirculation Flow Control System, pages 5. Quad Cities Recirculation System, page 3. ! Quad Cities Anticipated Transient Withou'c Scram, page 8. Quad Cities ATWS Learning Objectives, #9. 202001K201 202001K402- 202001K506 ..(KA's) ANSWER 6.09 (2.75)
- a. Outboard suction / isolation valve shuts (MO-1201-5).
RWCU pumps trip. Flow demineralizers go into hold. RWCU return isolation / discharge valve shuts (MO-1201-80) (3 required, 0.25 ea.)
- b. Low Reactor Water level Group 3 (OR +8",+/-2")
NRHX high outlet temperature 140 degrees (+/- 5 degrees) (2 required, 0.25 per signal, 0.25 per setpoint)
- c. Close fully (0.25). The flow control valve will shut when pressure upstream of the FCV decreases to lesc than 5 psig (0.75).
1 (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) 1
- y. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC Page 66 RESPONSIBILITIES (13%) q i
REFERENCE i Quad Cities RWCU system, pages 4, 5, 6, 8, 10. j
' Quad Cities RWCU Learning Objectives, #3, #5, #7.
204000A301 204000K404 ..(KA's) i ANSWER 6.10 (1.00) i REFERENCE Quad Cities Feed and Condensate system, page 14. Quad Cities Feed and Condensate system Learning Objective #4. 256000A201 ..(KA's) ANSWER 6.11 (1.50) I
- a. Scram inlet (0.25) and outlet valve are open (0.25) (for control rod j 34-27).
- b. Low nitrogen pressure (<950 psig) (0.25) or high accumulator water level (>37 cc past accumulator seal.) (0.25)
- c. Rod at odd-numbered reed switch (0.25) (or rod j not at even-numbered latch position) and no motion requested (0.25). !
REFERENCE 1 Quad Cities RMCS & RPIS page 4, QOA 900-5-G. ; Quad Cities RMCS & RPIS Learning Objectives, #6. j 214000A301 214000A402 ..(KA's) ANSWER 6.12 (1.00) b. 1 (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
.Pago 67 I
- 6. PLANT-SYSTEMS'(30%) AND PLANT-WIDE GENERIC RESPONSIBILITIES (13%)-
REFERENCE Quad Cities Reactor Vessel and Internals, page 14. Quad Cities Reactor Vessel and Internals Learning Objectives, #4, #9. 290002K403 ..(KA's) ANSWER 6.13 (3.00) (0.50) a) Neither Backup Scram Valve opens b) 125 VDC Bus (0.50) c)- 1. Remain Closed oc~ cwm c.t cd C " ,,
- 2. Close a ' p- 9~ -- &
- 3. Open ..
- 4. Remain closed (4. required, 0.50 ea.)
REFERENCE i LESSON PLAN ON 0500 RPS PG 8, 10, 12, AND 16. LEARNING OBJECTIVE NO. 3 AND 9. 201001G007(3.6/3.7) 201001K203(3.5/3.6) 201001K404(3.6/3.6) ; 201001K406(3.8/3.9) ...K/A (K/A VALUE) 201001K406 201001K404 201001K203 201001G007 ..(KA's)' ANSWER 6.14 (2.00)
/,2 a) Rod block (0.50) IRM ,16' will indicate 110 on range 6, rod block !
setpoint is 108. (0.50) b) No action (0.50) ! The downscale rod block is bypassed when the IRM range switch is on range 1. (0.50) 1 J (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
v
- 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC Page 68 RESPONSIBILITIES (13%)
REFERENCE LESSON PLAN ON IRM SYSTEM PG 10, 13, AND 20. LEARNING OBJECTIVE NO. 4 AND 9. 215003A304(3.5/3.5) 215003A401(3.3/3.3) 215003A403(3.6/3.4) 215003K401(3.7/3.7) ...K/A (K/A VALUE) 215003K401 215003A403 215003A401 215003A304 ..(KA's) ANSWER 6.15 (2.50)
- a. MO-7503A (Unit 2 Reactor Building suction damper) shuts (0.25)
MO-7503B (Unit 1 Reactor Building suction damper) opens (0.25)
- b. Low reactor water level, +8 inches High Reactor Building ventilation exh'aust radiation, 3mr/hr l High Refueling Floor Radiation, 100 mr/hr High Drywell radiation, 100 R/hr Selected SBGT train failure to start,.within 25 sec (or due to low flow 2000 scfm +/-500 scfm) l (4 required, 0.25 per signal, 0.25 per setpoint)
REFERENCE LESSON PLAN ON SBGT LIC 7500 PG 4, 14, an'd 18. LEARNING OBJECTIVE NO. 6 AND 9. 261000A211(3.2/3.3) 261000A303(3.0/2.9) 261000A402(3.1/3.1) 261000K101(3.4/3.6) 261000K401(3.7/3.8) ...K/A (K/A VALUE) 261000K401 261000K101 261000A402 261000A303 261000A211
..(KA's) l 4
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
- 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC Pago 69 RESPONSIBILITIES (13%)
ANSWER 6.16 (2.50) l a) 1. To preserve the integrity of the fuel cladding (La protect against excessive thermal heat flux that could perforate fuel cladding)
- 2. To preserve the integrity of the reactor coolant system (OR protect against excessive reactor pressure that could rupture the reactor I i
coolant boundary)
- 3. Minimize the energy which must be absorbed following a LOCA (and prevents criticality).
l (2 required, 0.50 ea.) { b) The mode switch in shutdown scram is bypassed after 2 seconds so ) that the operator can reset the scram with the mode switch in the shutdown position. (0.75) c) The operator is prevented from resetting the scram for 10 seconds i in order to ensure that all control rods have completed their l scram stroke into the reactor befor'e the scram is reset. (0.75) J REFERENCE LESSON PLAN 0500 ON RPS PG 2, 50, AND 56. , LEARNING OBJECTIVE NO. 3 AND 9. I 212000G004(4.2/4.3) 212000K408(4.2/4.2) 212000K412(3.9/4.1) I
...K/A (K/A VALUE) 212000K412 212000K408 212000G004 ..(KA's) I ANSWER 6.17 (1.00)
C. REFERENCE Quad Cities Operating Valve and Breaker Checklist, QAP 300-18, page 2. 294001K101(3.7/3.7) ...K/A (K/A VALUE) 294001K101 ..(KA's) (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
- 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC Page 70 RESPONSIBILITIES (13%)
ANSWER 6.18 (1.00) d. REFERENCE , i QAP 300-13 PG 1 AND 2. QAP 300-14 PG 1. 294001K102(3.9/4.5) ...K/A (K/A VALUE) 294001K102 ..(KA's) ANSWER 6.19 (1.00) b. REFERENCE QGA 300-13 PG'l AND 2. QGA 300-14 PG 3. 294001K102(3.9/4.5) ...K/A (K/A VALUE) 294001K102 ..(KA's) j l ; I ANSWER 6.20 (1.00) d. REFERENCE QEP 105-S1 PG 2. 294001A116(2.9/4.7) ...K/A (K/A VALUE) 294001A116 ..(KA's) l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
1
- 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC Page 71 RESPONSIBILITIES (13%) ,
ANSWER- 6.21 (1.00) Site Area (0.50) NARS Form 9C - 9F (0.50) REFERENCE QEP 100-T2 PG 1. QEP 200-T1 PG 7. 294001A116(2.9/4.7) ...K/A (K/A VALUE) 294001A116 ..(KA8s) ANSWER 6.22 (1.50) a) 2 am (0200) (0.50) (8 hour break between shifts) b) 22 hr (0.50) (requirement limiting the maximum number.of hours worked to 24 in a 48 hour period' c) 16 hr (0.50) (requirement limiting the maximum number of hours worked to 16 in a 24 hour period) REFERENCE QAP 300-1 PG 1 AND 2. 294001A103(2.7/3.7) 294001A109(3.3/4.2) ...K/A (K/A VALUE) 294001A109 294001A103 ..(KA's) ANSWER 6.23 (2.00)
- a. (1) Technical Staff Supervisor (0.50)
(2) Station Manager (0.50)
- b. 1. (1.00)
(***** CATEGORY 6 CONTINUED Oh :XT PAGE * * * * * )
- 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC Page 72
, RESPONSIBILITIES (13%)
REFERENCE QAP 1100-7 PG 1 AND 2. 294001A102(4.2/4.2) ...K/A (K/A VALUE) 294001A102 ..(KA's) ANSWER 6.24 (1.50) s a. safety man (0.50) ( 7~$ ', e / "/>c e )
- b. One (0.50)
- c. assign an individual at the' access point (0.50)
REFERENCE QAP 1120-6 PG 2. QAP 1900-9 PG 4. 294001K103(3.3/3.8) 294001K105(3.2/3.7) ...K/A (K/A VALUE) 294001K105 294001K103 ..(KA's) ANSWER 6.25 (1.00)
- 1. To determine that unpurged pockets do not exist (0.50)
- 2. To determine the general radiation conditions (0.50)
REFERENCE QAP 1150-3 PG 2. 294001K114(3.2/3.4) ...K/A (K/A VALUE) 294001K114 ..(KA's) l l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
v 1 6'- PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC Page.73 ' RESPONSIBILITIES (13%)- I j ANSWER 6.26 (2.00)
- 1. Any further degradation in the level of plant safety (or other worsening plant conditions.
- 2. Any change from one GSEP classification to another
- 3. Termination of a GSEP classification ,
- 4. Results_of ensuing evaluations or assessments of plant conditions- ]
- 5. Effectiveness of response or protective measures taken
- 6. Information related.to plant behavior that is not understood 1
(4 required, 0.50 ea.) l REFERENCE QAP 1290-1 PG 3. l 294001A116(2.9/4.7) ...K/A (K/A VALUE) 294001A116 ..(KA's) l l
\
.~
I 1 l
)
1 l l f (***** END OF CATEGORY 6 *****) (********** END OF EXAMINATION **********)
TEST CROSS REFERENCE Page 1-QUESTION VALUE REFERENCE 4.01 24.00 9000090 24.00 5.01 1.75- 9000045 5.02 1.75 9000046 5.03 2.00 9000047 5.04- 1.00 9000048 5.05 1.50 9000049 5.06 2.00 9000050 5.07 13.00 9000051 5.08 3.25 9000052 5.09 3.00 9000053 l 5.10 1.00 9000054 5.11 1.00 9000055 5.12 1.50 9000056 l 5.13 1.00 9000057 5.14 2.00 9000058 5.15 1.00 9000059 5.16 2.50 9000060 5.17 1.75 9000061 5.18 2.00 9000062 33.00 6.01 2.75 9000063 6.02 1.00 9000064 l 6.03 1.00 9000065 6.04 2.75 9000066 6.05 1.50 9000067 i 6.06 1.00 9000068 6.07 1.00 9000069 6.08 2.75 9000070 6.09 2.75 9000071 6.10 1.00 9000072 6.11 1.50 9000073 6.12 1.00 9000074 6.13 3.00 9000076 L 6.14 2.00 9000077 ! 6.15 2.50 9000078 6.16 2.50 9000079 6.17 1.00 9000075 6.18 1.00 9000080 6.19 1.00 9000081 6.20 1.00 9000082
- l. 6.21 1.00 9000083 l
6.22 1.50 9000084 l 6.23 2.00- 9000085 l 6.24 1.50 9000086 l 6.25' 1.00 9000087 6.26 2.00 9000088 43.00 _ _ _ . . _ _ _ _ _ _ . . . ~ _ _ . _ _ - - _ . , _ . _ _ _ . _ _ _ . _ _ _ _ _ . _ _ _ _ . . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . -
t r ^. m m ca LL r - G ,%t mcr. P e Y 1 l l 1 l l i ATTACHMENTS i Quad-Cities 1 I 198 /02/06 I SRO Examination l \ \ \ l l l l l 1 I l s l ' f'
.. L L b
- r
- y. ,, J -
v ; I b SELECTED TECHNICAL SPECIFICATIONS i FROM SECTION 3.5/4.5 l
]
i
QUAO-C8 TIES OPR-29 3.5/4.5 CORE AND' CONTAINMENT COOLING SYSTEMS s-LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS Applicability: Applicability: ! I I Applies to the operational status of.the Applies to periodic testing of the emer-gency cooling subsystems. l emergency cool!ng subsystems. Objective: Objective: ) To assure adequate cooling capability for To verify the operability of the core and heat removal in the event of a loss-of- containment cooling subsystems coolant accident or isolation from the normal reactor heat sink. SPECIFICATIONS I
)
l l A. Core Spray Subsystems and the LPCI A. Core Spray Subsystems and the LPCI j ' Mode of the RHR System Mode of the RHR System Surveillance of the core spray sub-systems and the LPCI mode of the RHR system shall be performed as follows:
- 1. Both core spray subsystems shall 1. Core Spray Subsystem Testing i l
i be operable whenever irradiated fuel is in the reactor vessel Item frequency I and prior to reactor startup l from a cold condition. a. Simulated auto- Each matic actuation refueling
- b. Flow rate test - After pump q core spray pumps maintenance j shall deliver at and every {
1 east 4500 gpm 3 months j against a system head correspond-ing to a reactor vessel pressure of 90 psig l c. Pump operability Once/ month ! I
- d. Motor-operated Once/ month valve 10118 3.5/4.5-1 Amendment No.
__ __________-_______________a
e QUAD-CITIES DPR-29
- e. Core spray w.
header a p instrumentation check Once/ day calibrate Once/3 months test once/3 months
- f. Logic system Each functional refueling test outage
- 2. From and after the date that one 2. When it is determined that one i
of the core spray subsystems is core spray subsystem is inoper-l made or found to be inoperable able, the operable core spray for any reason, continued reac- subsystem and the LPCI mode of tor operation is permissible the RHR system shall be l only during the succeeding 7 demonstrated to be operable days unless such subsystem is immediately. The operable core sooner made operable, provided spray subsystem shall be that during such 7 days all ac- demonstrated to be operable-tive components of the other daily thereafter, core spray subsystems and the LPCI mode of the RHR system and the diesel generators required for operation of such components if no external source of power 3. LPCI mode of the RHR system ; were available shall be operable. testing shall be as specified in 1 Specifications 4.5 A.I.a. b. c.
- 3. The LPCI mode of the RHR system d, and f, except that each LPCI shall be operable whenever division (two RHR pumps per irradiated fuel is in the division) shall deliver at least reactor vessel and prior to 9000 gpm against a system head corresponding to a reactor ves-l reactor startup from a cold i condition. sel pressure of 20 psig, with a minimum flow valve open.
- 4. When it is determined that one
- 4. From and after the date that one of the RHR pumps is inoperable, of the RHR pumps is made or the remaining active components found to be inoperable for any of the LPCI mode cf the RHR, reason,' continued reactor opera- containment cooling mode of the tion is permissible only during RHR, and both core spray the succeeding 30 days unless subsystems shall be demonstrated such pump is sooner made oper- to be operable immediately and able, provided that during sich the operable RHR pumps daily 30 days the remaining a:tive thereafter, components of the LPCI mode of the RHR, containment cooling 10118 3.5/4.5-2 Amendment No.
1
1 QUAD-CZT!ES DPR-29 mode of the RHR, all active compgnents of both core spray subsystems, and the diesel generators required for operatl'on of such components if no external source of power were 4 available shall be operable.
- 5. From and after the date that the 5. When it.is determined that the LPCI mode of the-RHR system is LPCI mode of the RHR system is i made or found to be inoperable inoperable, both core spray sub- '
for any reason, continued reac- systems, the containment cooling tor operation is permissible mode of the RHR shall be only-during the succeeding 7 demonstrated to be operable days unless it is sooner made immediately and daily thereafter, operable, provided that during such 7 days all active compo-nents of both core spray sub-systems, the containment cooling mode of the RHR (including two RHR pumps), and the diesel gen-erators required for operation . of such components if no exter-nal source of power were avall- , able shall be operable. I I'
- 6. If the requirements of Specifl-cation 3.5 A cannot be met, an orderly shutdown of the reactor shall be initiated, and the re- B. Containment Cooling Mode of the RHR actor shall be in the cold shut- System down condition within 24 hours.
Surveillance of the containment B. Containment Cooling Mode of the RHR cooling mode of the RHR system shall System be performed as follows: , 1
- 1. RHR service water subsystem '
testing: i Item Frequency l
- 1. a. Both loops of the containment cooling mode of a. Pump and valve Once/3 the RHR system, as defined operability months in the bases for Spe-cification 3.5.8, shall be operable whenever irradiated fuel is in the reactor vessel and prior to reactor startup from a cold condition.
( 10118 3.5/4.5-3 Amendment No.
QUAD-C8 TIES OPR-29
- 1. b. From the effective date of b. Flow rate test - After puno this amendment until July 1, each RHR service maintenance
+982, the "A" loop of the water pump shall and every containment cooling mode of deliver at least 3 months the. RHR system for each 3500 gpm against reactor may share the Unit 2 a pressure of 198 "A" and "B" RHR service psig water pumps using cross tie line 1/2-10124-16"-D. c. A logic system Each Consequently, the require- functional test refueling ments of Specifications outage 3.5.B.2 and 3.5.B.3 will impose the corresponding surveillance testing of equipment associated with both reactors if the shared RHR service water pump or pumps, or the cross tie i line, are made or found to l be inoperable.
- 2. From and after the date that one 2. When it is determined that one of the RHR service water pumps RHR service water pump is 1500-is made or found to be inoper- erable, the remaining components !
able for any reason, continued of that loop and the other con- I reactor operation is permissible tainment cooling loop of the RHR , only during the succeeding 30 system shall be demonstrated to ! days unless such pump is sooner be operable immediately and l made operable, provided that daily thereafter. during such 30 days all other active components of the con-tainment cooling mode of the RHR system are operable.
- 3. From and after the date that one 3. When one loop of the containment loop of the containment cooling cooling mode of the RHR system mode of the RHR system is made becomes inoperablo, the operable or found to be inoperable for loop shall be demonstrated to be any reason, continued reactor operable immediately, and daily operation is permissible only thereafter.
during the succeeding 7 days un- . I less such subsystem is sooner made operable, provided that all active components of the other loop of the containment cooling mode of the RHR system, both core spray subsystems, and both diesel generators required for operation of such components if no external source of power were available, shall be operable. 10118 3.5/4.5-4 Amendment No.
QUAD-CITIES DPR-29
- 4. Containment cooling spray loops 4. During each 5-year period, an are required to be operabie when air test shall be performed on ;
the reactor water temperature is the drywell spray headers and ' greater than 212' F and prior to nozzles and a water spray test i reactor startup from a cold con- performed on.the torus spray I dition. Continued reactor oper- header and nozzles. ; ation is permitted provided that a maximum of one drywell spray loop may.be inoperable for 30 days when the reactor water tem-perature is greater than 212' F.
- 5. If the requirements of 3.5.8 cannot be met, an orderly shut-down shall be initiated, and the reactor shall be in a cold shut-down condition within 24 hours.
C. HPCI Subsystem C. HPCI Subsystem Surveillance of HPCI subsystem shall be performed as follows:
- 1. The HPCI subsystem shall be 1. HPCI subsystem testing shall be l operable whenever the reactor as specified in Specifications pressure is greater than 90 4.5.A.I.a. b, c, and d, except psig, irradiated fuel is in the that the HPCI pump shall deliver reactor vessel, and prior to re- at least 5000 gpm against a sys- ,
actor startup from a cold condi- tem head corresponding to a re- l tion. actor vessel pressure of 1150 ) psig to 150 psig, and a logic system functional test shall be performed during each refueling outage.
- 2. From and after the date that the 2. When it is determined that the HPCI subsystem is made or found HPCI subsystem is inoperable, to inoperable for any reason, the LPCI mode of the RHR system, continued reactor operation is both core spray subsystems, the l permissible only during the suc- automatic pressure relief sub-I ceeding 7 days unless such sub- system, and the RCIC system system is sooner made operable, shall be demonstrated to be op-provided-that during such 7 days erable immediately. The RCIC all active components of the system snall be demonstrated to automatic pressure relief sub- be operable daily thereafter.
systems, the core spray sub- Daily demonstration of the auto-systems, LPCI mode of the RHR matic pressure relief subsystem system, and the RCIC system are operable. l l U Amendment No. l 1011B 3.5/4.5-5
QUAD-CIT 8ES DPR-29 l operability is not required pro '. p.
- vided that two feedwater pumps
-are operating at levels above 300 MWe: and one feedwater pump is operating as normally re- -i' quired with one additional feed-water pump operable at power levels less than 300 MWe.
- 3. If the requirements of Specift- l cation 3.5.C cannot be met, an orderly shutdown shall be initi-ated, and the reactor pressure i shall be reduced to 90 psig I within 24 hours. ,
l D. Automatic Pressure Relief Subsystems D. Automatic Pressure Relief Subsystems Surveillance of the automatic pres-sure relief subsystem shall be per-formed as follows:
- 1. The automatic pressure relief 1. The following surveillance shall subsystem shall be operable be carried'out on a six-month whenever the reactor pressure is surveillance interval:
greater than 90 psig, irradiated l fuel is in the reactor vessel a. With the reactor at pres- J l and prior to reactor startup sure each relief valve shall i from a cold condition. be manually opened. Relief I valve opening shall be i verified by a compen- sating turbine bypass valve or control valve closure. l
- 2. From and after the date that one 2. A logic system functional test of the five relief valves of the shall be performed each refuel-automatic pressure relief sub- ing outhge.
system is made or found to be inoperable when the reactor is , pressurized above 90 psig with l trradiated fuel in the reactor vessel, reactor operation is permissible only during the suc-ceeding 7 days unless repairs are made and provided that dur-ing such time the HPCI subsystem is oper&ble. l 10118 3.5/4.5-6 Amendment No. l 3
QUAO-CITIES OPR-29 ( 2. a. Plant operation shall be in accordance with 3.5.D.2 above.except that, for the . l current operating cycle 5, four of the five relief valves of the ADS are re-quired to be operable. In subsequent operating cy-cles, operation shall be in accordance with 3.5.D.2.
- 3. If the requirements of Specifi- 3. A simulated automatic initiation cation 3.5.0 cannot be met, an which opens all pilot valves orderly shutdown shall be initi- shall be performed each re-ated and the reactor pressure fueling outage, shall be reduced to 90 psig within 24 hours.
l
- 4. When it is determined that one relief valve of the automatic pressure relief subsystem is in-operable, the HPCI shall be demonstrated to be operable im-mediately and weekly thereafter'.
l E. Reactor Core Isolation Cooling System E. Reactor Core Isolation Cooling System l Surveillance of the RCIC system shall be performed as follows:
- 1. The RCIC system will be operable 1. RCIC system testing shall be as whenever the reactor pressure is specified in Specification greate.r. than 150 psig, irradi- 4.5.A.I.a. b, c, and d, except ated fuel is in the reactor ves- that the RCIC pump shall deliver sel,,and_ prior to startup from a at least 400 gpm against a sys-cold condition; tem head corresponding to a re-actor vessel pressure of 1150 psig to 150 psig, and a logic system functional test shall be run during each refueling outage.
- 2. From and after the date that the 2. When it is determined that the RCIC system is made or found to RCIC system is inoperable, the be inoperable for any reason, Fe' fHet system shall be demon-continued reactor operatica is strated to be operable immedi-permissible only during the suc- ately and daily thereafter.
ceeding 7 days unless such sys-tem is sooner made operable, provided that during such 7 days all active components of the HPCI system are cperable. 10118 3.5/4.5-7 Amendment No.
QUAO-C8T8ES DPR-29
- 3. If the requirements of Specif1-cation 3.5.BPT and 3.5.E.2 can! not be met, an orderly shutdown shall be initiated and the reactor pressure shall be reduced to 90 psig within 24 hours.
During the period November 6 through November 20, 1976 the unit may be started up with the RCIC inoperable provided that (1) the facility is not more than 7 days with the RCIC inop-erable and (2) all the HPCI system active components are demonstrated to be operable immediately after startup and daily thereafter. F. Minimum Core and Containment Cooling F. Minimum Core and Containment Cooling System Availability System Availability
- 1. Any combination of inoperable Surveillance requirements to. assure components in the core and con- that minimum core and containment tainment cooling systems shall cooling systems are available have not defeat the capability of the been specified in Specification 4.2.B.
remaining operable components to fulfill the core and containment cooling functions.
- 2. When irradiated fuel is in the reactor vessel and the reactor is in the cold shutdown condi-tion, all low-pressure core and containment cooling systems may be inoperable provided no work is being done which has the po-tential for draining the reactor vessel.
f 1011B 3.5/4.5-8 Amendment No. _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ w
I QUAO-CITfE5 OPR-29
- 3. When try,4diated fuel is trLthe
. reactor and the vessel head is '
removed, the suppression chamber may be dr'a~1ned completely and no more than one control rod drive I housing opened at any one time provided that the spent fuel pool gate is open and the fuel pool water level is maintained at a level of greater than 33 feet above the bottom of the I pool. Additionally, a minimum condensate storage reserve of ! 320,000 gallons shall be main-tained, no work shall be per- . l formed in he reactor vessel while a control rod drive housing is blanked following re- 1 moval of the control rod drive, and a special flange shall be available which can be used to blank an open housing in the , vent of a leak. , I
- 4. When irradiated fuel is in the l reactor and the vessel head is removed, work that has the po-tential for draining the vessel may be carried on with less than 112,200 ft3 of water in the suppression pool, provided that: (1) the total volume of water in the suppression pool, refueling cavity, and the fuel storage pool above the bottom of the fuel pool gate is greater than 112,200 ft3 ; (2) the fuel storage pool gate is removed; (3) the low-pressure core and containment cooling systems are operable; and (4) the automatic )'
mode of the drywell sump pumps I ! is disabled. l ' I i 3.5/4.5-9 Amendment No. j 10118
I
)
QEP 100-T2 PREDETERMINED PARS FOR CONTROL ROOM l l 1 QEP 200-T1 QUAD-CITIES EMERGENCY ACTION LEVELS I l j i l i l l l l I l l l l l l l u________ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _
f
~ PROCEDURE: QEP.100-1 QEP 100-T2 Revision I !
, . PREDETERMINE 0 PARS-FOR CONTROL ROOM August 1988 )
y 7.,?. I,,.f
, ]
NARS l 1- CLA55ff! CATION __
,, " CDN6}. TION FORM l 1 Ucu_r - ,
Unusual Event All conditions' 9A l l Alert (1) Potential Release or No Release 9A m :, (2) Release Occurring , 9B ; i i Site Area (1) Potential Release.or No Release 98 (2) Release Occurring 9C thru. l 9F General Emergency Select.a choice from Section A g from - Section B on the following page. Complete the NARS form using the selection with the highest severity level (1 55 highest). ; i l I a 1 l l APPROVED aun 041988
. . o.c o s.R 14/0357e i
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