ML20247B122
| ML20247B122 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 07/17/1989 |
| From: | Burdick T, Reidinger T, Shembarger K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20247B108 | List: |
| References | |
| 50-282-OL-89-01, 50-282-OL-89-1, 50-306-OL-89-01, 50-306-OL-89-1, NUDOCS 8907240041 | |
| Download: ML20247B122 (175) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION III Reports No. 50-282/0L-89-01; 50-306/0L-86-01 Docket Nos. 50-282; 50-306 Licenses No. DPR-42; DPR-60 Licensee: Northern States Power Company 414 Nicollet Mall Minneapolis, MN 55401 Facility Name: Prairie Island Examination Administered At: Prairie Island Examination Conducted: Senior Reactor Operator and Reactor Operator RIII Examiners: 1 T. Reidinger '
71 Dhte
/N IK. Shembarger/
h 8L I)di Date ' '
Approved By: QWh W q ~7!Oh Thomas M. Burdick, Chief Date ' /
Operator Licensing Section Examination Summary Examination administered on June 13-15, 1989 (Reports No. 50-282/0L-89-01; No. 50-306/0L-89-01).
Areas Inspected: Written and/or operating exams were administered to five reactor operators and one senior reactor operator.
Results: All candidates passed the examinations, gg7240043g9o737 ,
V ADOCK 05000282 '
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REPORT DETAILS
- 1. Examiners N. Maguire-Moffitt, PNL S. Carrick, PNL K. Shembarger, NRC T. Reidinger, NRC i D. Shepard, NRC 1
- 2. Exit Meeting An exit meeting was held on June 16, 1989, with facility management and training str.ff representatives.
NRC representatives in attendance were:
T. Reidinger, Chief Examiner D. Shepard, Examiner K. Shembarger, Examiner N. Maguire-Moffitt, PNL Contractor Examiner S. Carrick, PNL Contractor Examiner J. Hard, Senior Resident Inspector T. O'Connor, Resident Inspector Facility Represent:tives in attendance were:-
l l L. Waldinger, Manager, Production Training i D. Westphal, Instructor I
D. Reynolds, Operations Training Supervisor J. Jensen, Instructor j M. Gardizinski, Instructor T. Amundson, General Superintendent of Training D. Mendell, General Superintendent of Engineering and Radiation Protection M. Sellman, General Superintendent of Plant Operations The examinars observed the following generic strengths during tne administration of the operating examinations: (1) Communications between crew members durin I excellent, and (2)g use the of simulator portion procedures of the by tne examination crews during thewere simulator ,
portion of the examination were very good. No generic weaknesses were !
observed.
l l The facility review of both examinations in the region was comprehensive l
and thorough. To aid in the development of future examinations, an alarm response guidt index or annunciator layout diagram and the safety manual should be included in the submittal of the examination reference material.
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The examiners and. facility training representatives identified simulator-fidelity problems during scenario validation and during the administration-of the simulator portion of the .perating o examination.-lA list of the a specific weaknesses ~was given to the facility training representatives. ;j At the initiation of the written examination,.the. facility was provided a- .
copy of the examination and answer key for both the SR0 and R0 ;
examinations. .The facility provided written commente-~concerning'the:
examinations to the NRC prior to th exit meeting on June.16, 1989.
The following paragraphs contain the facility. comment.s concerning the'- .
examinations, followed by the NRC comnents, i s'
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REACTOR OPERATORS EXAMINATION QUESTION 1.03 (1.00)
During a reactor startup from an initial Keff of 0.88, the first reactivity addition caused count rate to increase from 10 counts per second to 15 counts per second. The second reactivity addition caused count rate to increase from 15 counts per second to 30 counts per second.
Which one (1) of the following statements describes the relationship between the first and second reactivity additions?
(a.) The first reactivity addition was the larger of the two.
(b.) The second reactivity addition was the larger of the two.
(c.) The first and second reactivity additions were approximately equal.
(d.) There is not enough data to datermine the relationship.
ANSWER 1.03 (1.00)
(c.)
COMMENT 1.03 The correct answer is (a.)
NRC RESPONSE The answer key was revised to reflect (a.) as the only correct answer.
QUESTION 1.14 (1.5) (Part b.)
- b. STATE THREE (3) types of devices (no moving parts) that use this principle.
ANSWER 1.14 (1.50) 'Part b.)
- b. 1. orifice
- 2. venturi
- 3. flow nozzle
- 4. delta-P across pipe bend I i
Any three (3) [+0.33] each i COMMENT 1.14 Add annubar to the list of stationary flow devices a
HRC Response l
l Annubar was added to list of stationary flow devices as an additional l correct answer. l 1
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l QUESTION 1.20 (1.00)
WHICH ONE (1) of the following statements describes how the steam generator pressure indication will compare to actual steam generator pressure if l containment pressure is increased to abnormally high pressure.
(a.) Indicated steam generator pressure will be HIGHER than actual steam generator pressure.
(b.) Indicated steam gcuerator pressure will be LOWER than actual steam generator pressure.
(c.) Indicated steam generator pressure will be EQUAL to actual steam generator pressure.
(d.) Indicated steam generator pressure will fail as is.
ANSWER 1.20 (1.00)
(b.)
COMMENT 1.20 The correct answer is (c.) i NRC Response The answer key was revised to reflect (c.) as the only correct answer.
QUESTION 3.05 (2.00) (Part a.)
Assume Unit 1 is at 100% power. Power range nuclem instrumentation channel N-41 fails high.
- a. WHAT FOUR (4) protection / control circuits are affected by this channt.1 failure?
ANSWER 3.05 (2.00) (Part a.)
- a. 1. OT delta-T Lv3.25]
- 2. OP delta-T [+0.25]
- 3. power range h+gh flux trip [+0.25]
- 4. power range high flux rod stop [+0.25]
COMMENT 3.05 The following should be included as additional correct answers:
- 1. Rod control (power mismatch circuit)
- 2. Low flux trip i
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- 3. Positive rate ~ trip
- 4. _ Negative rate" trip
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..NRC Response The' answer key was revised as follows:- 1
'a. '1. OT' delta-TJ
- 2. -
OP: delta-T,
-3.. power-rangehighflux" trip-
- 4. power range high. flux rod stop' 5.; rod control (power; mismatch circuit)-
6.L low flux trip-. _
.l 7.9 Positive rate trip
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- 8. -Negative rate-trip _ '
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.[any 4;at 0.25 each] 'l i
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. - - - . . - _ - - - _:...-.--...!1 -- ----_.__--__LL_-- U_
SENIOR REACTOR OPERATOR'S EXAMINATION QUESTION 4.18 (1.00)
Which ONE of the following statements describes how the steam generator pressure indication will compare to actual steam generator pressure if containment pressure increased to an abnormally high pressure during a LOCA: (1.0)
- a. Indicated steam generator pressure will be HIGHER than actual steam.
generator pressure.
- b. Indicated steam generator pressure will be LOWER than actual steam generator pressure.
- c. Indicated steam generator pressure will be EQUAL T0 actual steam generator pressure,
- d. Indicated steam generator pressure will fail an is.
ANSWER 4.18 (1.00)
- b. Indicated steam generator pressure will be lower than actual steam generator pressure.
COMMENT 4.18 The correct response is "c. Indicated steam generator pressure will be EQUAL T0 actual steam generator pressure", since the transmitters are not located inside containment.
HRC Response The answer key was revised to reflect c. as the only correct answer.
QUESTION 5.16 (2.00) (Part a.)
- a. IDENTIFY the Emergency Operating Procedures that would be used during the RECOVERY and C00LDOWN from a loss of all offsite power. (Assume no further complications occur). (1.0)
- ANSWER 5.16 (2.00) (Part a.3)
(Procedure Number or Procedure Title required for full credit)
- a. 3. IES-0.3, Natural Circulation Cooldown with CRDM Fans. (0.33) i 9
COMMENT 5.16 The CRDM fans would not be operable due to the loss of offsite power.
Therefore, the correct answer should be: IES-0.3 Natural Circulation Cooldown without CRDM fans. )
NRC Response I The answer key was amended to reflect the revised answer.
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. 1 SIMULATION FACILITY REPORT Facility Licensee: Preirie Island i Facility Licensee Docket No. 50-282 l Operating Tests Administered At: Prairie Island l l
i During the conduct of the simulator portion of the operating tests, the j following items were observed (if none, so state): ;
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ITEM DESCRIPTION i I
- 1. 50 MW swing in turbine output malfunction was not functional during the ;
examination. ]
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- 2. ATWS and loss of RCS flow event resulted in depressurization of RCS and i automatic SI actuation. 1
- 3. Spurious activation of a'inuciators during examination. q l
- 4. Generator H, temperature sensor TEIS047 low failure malfunction didn't i result in a high temperature indication.
- 5. During a steam generato? tube leak event, the quick leak calculation j computer remained frozen at 32 gpm throughout the event.
- 6. Reuiation monitor indication for steam generator tube leak and leakage into l component cooling water system appeared not to realistically model actual i tube leak radiation and indications.
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MASTER COP.Y l l
- 0. S. NUCLEAR REGULATORY COMMISSION i SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: ERAIRIE ISLAND REACTOR TYPE: PWR-WEC2 l 1
DATE ADMINISTERED: 89/06/12 CANDIDATE'S NAME:
l INSTRUCTIONS 'LO CAN,DIDATE:
Use separate paper for the answers. Write answers on one side only.
Staple question aheet on top of the answer sheets. Points fo each question ore indicated in parentheses after the question. The passing )
grade requires at least 70% in each category and a final grade of at !
j least 80%. Examination papers will be picked up six (6) hours after ths examination starts. . l
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY
_24.00 24.00 4. REACTOR PRINCIPLES (7%) i THERMODYNAMICS (7%) AND i COMPONENTS (10%) (FUNDAMENTALS EXAM)
EMERGENCY N G ABNORMAL PLANT l 33.00 33.00 5.
EVOLUTIONS s33%)
43.00 43.00 6. PLANT SYSTEMS (30%) AND >
PLANT-WIDE GENERTC RESPONSIBILITIES (13%)
100 . O ___ % TOTALS FINAL GRADE All work done on this examination is my own. I have neither given nor received aid.
Candidate's Signature MXSTER COPY
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During administration of this examination, the following rules apply:
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- 1. Cheating on the examination means an automatit denial of your I application and could result in more severe penalties.
- 2. After the examination has been completed, you must sign the .
statement on the cover sheet. indicating that the work is your own and !
j you have not received or given assistance in completing the examination. This must be done after you complete the examination.
- 3. Restroom trips are to be limited and only one applicant 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.
- 4. Use blank ink or dark pencil ONLY to facilitate legible reproductions.
- 5. Print your name in the blank provided in the upper right-hank corner of the examination cover sheet.
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- 6. If you need more space to write your answers other than that provided j on the examination sheet, use a separate sheet of the paper provided j and insert it directly after the specific question. . include the j question number on the separate sheet. DO NOT WRITE ON THE BACK SIDE OF THE EXAMINATION QUESTION PAGE.
- 7. . Print your name or initials in the upper right hand corner of each page of the examination and on any additional sheets of paper used. ]
- 8. Use abbreviations only if they are commonly used in the facility literature. Avoid using symbols such as > or < signs to avoid a simple transposition error resulting in an incorrect answer.
- 9. The point value for each question is indicated in parentheses after the question. Since there is only one question per page, the amount of blank space on each page is NOT an indication of the depth of answer required.
- 10. Show all calculations, methods, or assumptions used to obtain an answer. 1 11e Partial credit may be given. Therefore, answer all parts of the ;
l question and do not leave any answer blank. l
- 12. Severai questions on the exams are multiple choice. Only the !
circled answer will be graded. If more than one answer is circled, ;
no credit will be given. . Additional information provided by the i
candidate on the multiple choice questions will not be evtiuated.
- 13. Proportional. grading will be applied. Any additional wrong information that is provided may count against you.
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- 14. If the intent of a question is unclear, ask questions of the examiner only. !
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- 15. When turning in your examination, assemble the completed examination with examination questions, answers and examination aids. In cdditio's, turn in all scrap paper. 1
- 16. To past-the examination, you must achieve an overall grade of 80% or greater and at least 70% in each category. I'
- 17. There is a time limit of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for completion of the examination.
- 18. When you are done and have turned in your examination, leave the 1 examination area. If you are found in this area while the examination l
is still in progress, your license may be denied or revoked.
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Pega 1 of 5
- DOIG_SSEEI BEOCIQB UEQBY_EQ8dWLOS:
SUR(t)
P = p et/T P = P 10 p .
I'y_ID v---------- SUR = 26.06/v
__________10 fissions /sec 3.12 x 10 .
2 1 eff 1
-(B L th 2' " ~~ # ~~~~~~
p . _____________ . . y 1+ y th 2 2 1+ (B L th I 2 K- 1 -
-(B Lf 3 A = --<- --
P =e f K op = In ' ""
p , ,-EN3C1,9,3tlI, K initial 7I - p
=C **~ ~~~~
C 3
(1-K,,qg) 2 - eff2' p 1 C l'
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ALj 2 + AL th 2 a = - -- + - -- - B (- -- -----)
i p at at At f at l.
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g _______________________ ;
1 In H2 /R 3 in R3 /R 2 O=UA (AT*) - + -------- + - '
K K 2
K' 3 <
G=mc (AT)
P $ . a & A R* j e
0 (h n _ _Le_:_0 _eut y . __te___h_ gut _)teg1__ j 1
(h -h Q
in in out) ideal )
l actual 1 1 22 T
W T 2 .-
supplied 1 i
& = pAV pAV g g 3 = p2^2 2 j 3
E = KA J AP p E = KA J5 = KA AT E = KA op 4 Ae ,
u nc Q 1 AT . A_T___( i n_)_ __ _ _ A_ _T __ _ _( o u _t _) . . _f_E_ t_h____9 '
m B.Bx10 1 AT (in) 1n ( ------->
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K K K a b n 9
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CENIBIEUQ0L_EUdE LOWS:
Ng Ag (N g ) Hg (N g ) Pg
-_ __ --__ . __ -__.g = -_
N2 (N ) H (N P 2 ,
2 2 2 2 BSDIGIlON_8SD_CHE01SI61.EQBdWL8S:
R/h. = 6CE/L I, = 10 e CVg g =C 22 n -Gt G = Ullutlgn_ Bate I=I g ILL -
C=CO*
Volume JO A=A O ,- At A = AN CONYEGSIONS:
l'gm/cm = 62.4 lbm/ft Density of water (20 C) = 62.4 lbm/ft 1 gal = B.345 lbm 23 1ft = 7.48 gal Avogadro's Number = 6.023 x 10 1 gal = 3.78 liters Heat of Vapor (H O) = 970 Btu /lbm 2
1 lbm = 454 grams Heat _of Fuzion (ICE) = 144 Btu /lbm
~
e = 2.72 1 AMU = 1.66 x 10 grams !
w = 3.14159 Mass of Neutron = 1.008665 AMU 1 KW = 738 ft-lbf/sec Mass of Proton = 1.007277 AMU 1 KW a 3413 Btu /hr Mass of Electron = 0.000549 AMU 1 HP = 550 ft-lbf/sec One atmosphere = 14.7 psia = 29.92 in. Hg 1 HP = .746 kW 'F = 9/5 *C + 32 l 1 HP = 2545 Btu /hr *C = 5/9 (
- F - 32) 1 Btu = 778 ft-1bf 'R nF + 460
-16 *K = *C + 273 1 NEV = 1.54 x 10 Stu ,
~I h = 4.13 x 10 M-sec 0 *
( 1 W = 3.12 x 10 fissions /sec gg = 32.2 lbm-ft/lbf-sec c = 931 MEV/ AMU 0
'1 inch = 2.54 cm C=3x 10 m/sec
-8 R r = 0.1714 x 10 Btu /hr ft
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Page 4 off5 .
. 'D0Ie_SUEEI ;
. EVE 885E_IBEBd8L_CQWUUCIIVIIY_1El .
datati.al K- '
Cork' 0.025 Fiber Insulating Board 0.028 .j i
M:ple or Dak Wood i O.096 Cuilding Brick -0.4 '
Window Glass O.45 0.79 .;
- Cancrete '
'1% Carbon. Steel 25.00
~1% Chrome Steel 35.00 l
Aluminum .118.00 Copper 223.00 Silver 235.00
'Wster> (20 psi a, 200 degrees'.F) 0.392 Steam (1000 psia, 550 degrees F) ~0.046 l Urenium DioxAde. 1.15 i Haltuu 0.135 Zircaloy 10.0 <
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51SCELLANEQUS_INEQBdGIl0N:
E = sc l
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KE = 1/2 mv '
PE = mgh V, = VO + at ,
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Geometric Dbject Area Volume P Triangle A = 1/2 bh /////////////////
Square A m S* /////////////////
Rectangle A=LxW /////////////////
Circle A = pr* /////////////////
Rectangular Solid A = 2(LxW + LxH + WxH) V=LxWxH t
Right* Circular Cylinder ~A =J (2 wr*)h'+-2(wr#) V = wr#h 1
2 Sphere A = 4 wr *
'V~=~4/3'(wr#)
-Cube ///////////////////////////// V=S #
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515CELLONEQUS_1NEQ3d8110N_1coattauedi:
10 CFR 20 Appendix B Table 1 Table II Gamma Energy Col I Col II Col I Col II MEV per Air Water Air Water Mntsrial Half-Life Disintegration uc/ml uc/ml uc/ml uc/ml
-6 ~0 '
Ar-41 1.84 h 1.3 Sub 2x10 ----- 4x10 ------ .;
~# ~3 ~
~3 Co-60 5.27 y 2.5 5 3x10 1x10 1x10 " 5x10 .
~Y ~D ~l ~#
1-131 B.04 d 0.36 S' 9x10 6x10 1x10 3x10
~D ~#
Kr-85 10.72 y 10.04 Sub 1x10 ----- 3x10 ------
~# ~3 ~8' ~
Ni-65 2.52 h .O.59 S .9x10 4x10 3x10 1x10 "
-22 ~ ~l" ~6 Pu-239 2.41x10* y 0.000 S 2x10 1x10 *' 6x10 5x10
~ ~ -Il ~#
Sr-90 29 y' ----- S 1x10 1x10 " 3x10 3x10 ,
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Xa-135 9.09 h 0.25 Sub 4x10 ' ----- 1x10 ------
Any single radionuclides with T > 2 he ~
-10 -6 3x10 '
~
which does not decay by alpha be..' 9x10 1x10 3x10 spontaneous fission 2
Neutron Energy (MEV) Neutrons per cm Average flux to deliver equivalant to 1 rem 100 mrom in 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> ;
thermal 970x10'6 670 0.02 400x10 260 (neutrons) 6 30 ----------
- 0.5 43x10 10 24x10 6
- 37 c,2 u ,,c Linear Absorption Coefficients p (cm-A)
Energy (MEV) Water Concrete Iron Lead 0.5- 0.090 0.21 0.63 1.7 1.0 .O.067 0.15 0.44 0.77 1.5 0.057 0.13 0.40 0.57 2.0 0.048 0.11 0.33 0.51 2.5 0.042 0.097 -0.31 0.49 3.0 0.038 0.089 0.30 0.47
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- 4. REAgTOR PRINCLE[iES (7%) THERH0 DYNAMICS. Page- 4 l
.L7%) AND COMPONENIS (10%) (FUNDAMENTALS EXAM) l QUESTION 4.01 (1.00)
Which ONE of the following statements is CORRECT with regard to I criticality: (1.0)
- n. Critical rod height does NOT depend on how fast control rods are withdrawn.
- b. Critical rod height dictates the reactor power level when criticatily is achieved.
- c. The SLOWER the approach to criticality, the LOWER'the reactor power level will be when reaching criticality.
- d. Once a reactor startup is commenced, criticality.should be anticipated only after permissive P-6 is received.
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- 4. _ REACTOR _ PRINCIPLES (7%1 THERMODYNAMICS Page' 5 )
-(7%) AND COMPONENTE l10%) (FUNJ2bMENTALS EXAM 1 i
QUESTION 4.02 (2.00) e I
If'rcactor power increases from 1000 cps to 5000 cys in'30 seconds, <.nat is )
the Start Up Rate? SHOW ALL WORK (2.0)
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- 4. REACIOR PRINCIPLES (7%) THERMODYNAMICS Page 6 l (7%) AND COMPONENTS (10%) { FUNDAMENTALS EXAtil i
'j QUESTION 4.03 (1.00) ;
Which ONE of the following correctly represents.the contributors to the total power defect at BOL in INCREASING order of magnitude: (1.0)
- c. Void, Doppler, MTC .
- b. HTC, Void, Doppler-
- c. Void, MTC, Doppler .)
- d. HTC, Doppler, Void i
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'4. M CTOR PRINCIPLES R %) THERMODYNAMICS Page 7 (7%) AND COhPONENTS (10%) (FUNDAMENTALS EXAM) 9UISTION 4.04 (1.00)
Which ONE of the following statements is CORRECT concerning Xe-135- !
9roduction and removal: (1.0)- -
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- a. At full power, with equilibrium conditions, approximately one half of the xenon is produced by iodine decay and the other half-is produced directly from fission.
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- b. When a reactor is shutdown after 120 days of operation at 100% power, xenon burnup effectively' stops while the decay of iodine continues.
Therefore, the xenon concentration starts to increase.
- c. The production of xenon from iodine decay continues for up to approximately 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br /> after a reactor shutdown. Because all production of xenon has ceased at this time, xenon concentration- !
reaches its minimum level in.the reactor core. 1 I
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- 4. RE6QTOR EBINQlELES (7%1 THERH_QDYNAMlqS Page 8 I' il% LAND COMPONENIS (10%) LFUNDAMEHIALS EXAM 1 l
QUESTION 4.05 (1.00) )i Which ONE of the following will result in an INCREASE in the differential l boron worth: (1.0)
- a. Boron concentration INCREASES
- b. Moderator temperature INCREASES
- c. Fission product concentration INCREASES
- d. Core age INCREASES
- - - - - ----~~--~m- ._ m ..s
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~4. REA_GIOR PRIMGLPLES [7%) THERtjoDYNAfflGS (7%) AND COMPONENTS (10%) (FUNDAMENTALS EXAM), !
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QUESTION 4.06 (1.00)
Which ONE of the following is the MOST accurate definition of the term J roactivity: (1.0) f
-a. The rate of change of xeactor power in neutrons per second. .
l .b. The' ratio between the populations of two successive neutron generations. ,
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- c. The difference between critical and all control rods withdrawn for {
a given core. .q
- d. The measure of a reactor's departure from criticality, f I
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- 4. REA_QTOR EglNCIPLES__17%) THEBUQDYNAM19S Page 10 (7%) AND COMPONENTS (10%) (FUNDAMENTALS EXAM) :
i QUESTION 4.07 (1.00) ,
M2tch the PRIMARY heat transfer mechanism from COLUMN E that is taking i place in each of the conditions in COLUMN A. (1.0) ;
i COLUMN B COLUMN A
- 1. Nucleate boiling on_the cladding a. Convection surface of the fuel assembly.
- b. Radiation
- 2. Accident condition in which high quality. steam is passing through c. Conduction the coolant channels.
- 3. Beat from fission across a fuel pellet.
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- 4. REACTOR _EglHCIPLES (7%) THERM _QQYNAlilCS Page 11 ;
(7%1 AND COMPONENTS (10%)_LEUNDAMENTALS E M 1 QUESTION 4.08 (1.00)
Fill in the blanks concerning Reactor Vessel Integrity: (Each blank j i
m:y require more than one word). ( 1. 0.)
- o. Neutron embrittlement of the reactor core cau.ses to )
increase. This means that brittle fracture can occur at higher l temperatures. (0.5) i I
- b. Allowable pressures inside the reactor vessel are most. limiting during (HEAT-UP, COOLDOWN, or STEADY STATE) i (0,5) j operations.
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1 BEAGIQ8 PRINGLPLES_{.7%) THERHQDYHAMLQS Page 12 (7%) &HD_QQMPONENI U 10%) (FUNDAMENTALS EXAH_1 QUESTION '. 09 (1.00)
Which ONE of the'following statements is correct:
The 2200 degree F maximum peak cladding temperature limit 1.s used )
.bre uso: (1.0) . l 1
l a. it is 500 degrees F below the fuel cladding melting point.
- b. any clad temperature higher than this correlates to a fuel centerline temperature at the fuel's melting point.
- c. a zircalloy-water reaction is accelerated at temperatures above 2200 degrees F.
- d. the cladding becomes weaker, because of a zirconium phase change at temperatures above 2200 degrees F (e.g. zirconium goes from a close packed hexagoual structure to one that is body-centered cubic).
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_t__EEACTOR_PJINCIPLES 1731 THERMODYN_AtjICS Page 13 I (7%) At!D COMPONENId_110%) (FUNDAMENTALS EXAM 1 QUESTION 4.10 (1.00)
Which ONE of the following statements is CORRECT regarding the Departure from Nucleate Boiling Ratio (DNBR): (1.0)
- c. As Tavg decreases, the DNBR decreases.
- b. If rods are withdrawn without changing turbine load, the DNBR increases.
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- c. As pressurizer pressure increases, the DNBR increases.
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- d. As reactor power increases, the DNBR increases.
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! 4. REACTOR PRINCLELES (7%1 THERM _QDMAMICS Page 14 _q L 17%) AND cot 020NENIS (10%) (FUNDAMENTALS EMtil l
I QUESTION 4.11 (1.00)
.Which ONE of the following sets of conilitions ensures the. hot channel l factor limits in Technical Specifications will not be exceeded:- (1.0)
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- a. 1. Control rods in a single bank movs together with no individual rod insertion differing by more than 15 inches from the bank demand position.
- 2. Control banks are sequenced with overlapping banks.
- 3. Control rods are fully withdrawn while operating at 100% power. i l
- b. 1. Reactor thermal power limits.are observed.
- 2. Control rods are fully withdrawn while operating at 100% power.
- 3. Axial power distribution control procedures are observed.
- c. 1. Control bank insertion limits are not violated. ;
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- 2. Reactor thermal power limits are observed.
- 3. Control banks are sequenced with overlapping banks. l
- d. 1. Cor, trol rods in a single bank move together with no individual-r9d insertion differing Ly more than 15 inches from the bank )
demand position. 1
- 2. Cantrol bank insertion limits are not violated.
- 3. Ax;sl power distribution control procedures are observed. j i
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- 4. REAGTQR PRINglP_l,ES (7%) THE8HODHAMICS Page 15 (7%) AND_ COMPONENTS (10%) (FUNDAMENTALS EXAH1 QUESTION 4.12 (2.00)
The following questions concern the pressurizer level transmitters:
- a. State the basis for instelling sealed reference legs on the hot calibrated pressurizer level transmitters. (1. 0 )
- b. Linear motion of the pressurizer level transmitters is sensed by a silicon strain gage beam. What determines the magnitude of the linear motion? (Include in your answer the transmitter design feature used in determining the magnitude of linear motion).
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CATEGORY 4 CONTINUED ON NEXT PAGE *****)
ju_ REACTOR PRINCIPLES (7%) THERMODYNAMICS Page.16 (7%) AND COMPONENTS (10%) (FUNDAMENTALS EXAM) i QUESTION 4.13 (1.00)
I Which ONE of the following effects will occur as a result of increased fouling of the main condenser tube bundles while at 100% power: (Assume circulating water flow and steam flow rates remain constant) (1.0)
- a. Condenser heat rejection will increase
- b. Circulating water outlet temperature will decrease
- c. Condensate depression will increase I
- d. Condenser hotwell temperature will-occrease 1
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- 4. REAQTOR PRINgIPLES (7%1 THERMODYNAMICS Page 17 11%1_6ND COMPONENTS (10%) (FUNDAMENTALS EXAM)
QUESTION 4.14 (1.00)
Which ONE of the following statements DEST describes what happens to a fluid as it passes through a venturi: (1.0)
- a. Pressure remains constant, and the velocity increases as the diameter of the venturi decreases.
- b. Pressure increases, and the velocity decreases as the diameter of the venturi decreases.
- c. Pressure increases, and the velocity remains constant as the diameter of the venturi increases.
- d. Pressure increases, and the velocity decreases as the diameter of the venturi increases.
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- 4. REACTOR PRINCIPLES (7%) THERMODYNAMICS' 'Page 18 '
lZMl_899_G9dE9NENIE_11931_1EUND8tENIBLE_EXBd1 ~ .
f DUESTION 4.15 (1.00)
Which'ONE of the f ollowing statements about pump ACTUAL Net Positive Suction Head (NPSH) is CORRECT: (1.0) y
- a. When-a pump is started, the NPSH will decrease by the amount of the*
pressure drop in the suction piping.
- b. NPSH is the amount by which the saturation pressure is greater .than the !
suction pressure for the water being pumped. ,
- c. NPSH is essential for operation of centrifugal pumps, but_not for positive displacement pumps,
- d. NPSH can be calculated by subtracting the suction pressure f rom the- i 4
discharge pressure.
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l Fage 19 4 __BE69I98_EB1991ELEE_1231_IMEBd9920601gg IZ31_6BD_995E9BENIE_119%) 3EUUD6DESIBLS_gE8dl DUESTION- 4.16 (1.00)
Whi ch DNE of the f ollowing statements is CORRECT concerning steam generator.
oper a ti on: (1.0)
- a. Once the normal operating condition f or' a s' team generator has been established, UA (the overall coefficient of heat. transfer x heat transfer surface area) no longer i s a fixed value.
- b. _If the Delta T across a. steam generator is not constant, then the median (average) temperature i s used to accurately calculate the heat transfer rate.
- c. The heat removal rate for a steam generator will increase on1y when BOTH fluid flow rates through the steam generator increase.
- d. One advantage of recirculation flow in a steam generator is that as the level in the downcomer increases, the delta P driving head increases.
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dz__BESGI98_ESINGlELE5_l231_THEBUppVNBdJgg Page 20 IZZ1_SNp_GQUEQUENIg_J1931_JEgypBMEylBLg_gfBdl l
'DUESTION 4.17 (1.00)
Which DNE of the f ollowing MOST accurately describes the steam relief -
capacity of the Main Steam System: (1.0) \
a.. The code saf eties MUST be combined with the atmospheric dump valves to have a capacity of 100%, with the relief settings from'1005 psig to 1020 psig.
- b. The code safeties MUST be combined with the atmospheric dump valves to i I
have a capacity of 100%, with the relief settings from.1075 psig to. I 1129 psig.
- c. The code safeties DO NOT have to be combined with the atmospheric dump valves to have a capacity of 100%, with relief settings from 2005 psig to 1020 psig.
- d. The code safeties DO NOT have to be combined with the atmospheric dump valves to have a capacity of 100%, with relief settings from 1075 psig I
to 1129 psig.
- - - - - - = - _ - _ _ _ _ _ _
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'S___SEGGI98_EBIN91ELES_1231_IBEBdDDYN6bigg JZ31_6ND_G90E9BENIS_11931_lEUN98dENIBLS_EEedl DUESTION 4.18 (1.00)
Which ONE of the following statements describes how the steam generator pressure indication will compare to' actual steam generator pressure if-containment pressure is increased to an abnormally high pressure during a LOCA: (1.0)
- a. Indicated steam generator pressure will be HIGHER than actual steam.
generator pressure,
- b. Indicated steam generator pressure will be LOWER than actual steam generator pressure.
- c. Indicated steam generator pressure will- he EDUAL TO actual steam generator pressure.
- d. Indicated steam generator pressure will. fail as is.
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Page 22 dz__EE8GI98_EB19GIELEE_1ZLl_IHERMQDyN@dlGS' 1ZKl_6N9_G9dE9EEGIE_11911_1Eugo6 DEN 166E_EEGdl DUESTION 4.19 (1.00)
Which ONE of the f ollowing statements correctly. describes why density compensation of the main steam line flow measurement is necessary: (1.0)
- a. Differential pressure across the flow venturi in the steam line is proportional to the volumetric flow rate. Volumetric flow rate is compensated with the fluid density to provide mass flow rate.
- b. Differential pressure across the flow venturi in the steam line is inversely proportional to the volumetric flow rate. Volumetric flow rate is compensated with the fluid temperature to provide mass . flow rate.
- c. The temperature of the steam lines is proportional to the volumetric flow rate. Volumetric flow rate is compensated with the fluid temperature to provide mass flow rate.
- d. The temperature of the steam lines is inversely proportional. to the volumetric flow rate. Volumetric flow rate is compensated with the
- fluid density to provide mass flow rate.
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.St__BEBGI9B_EBIN91ELEg_JZZ.l_IBEBd9pyN6digg 12Z1_ BUD _99dE9 BENTS . J IQZ.)- (EUbpBdENI669_EZ801 DUESTION 4,20 (1.00)
Which DNE of the following would result in the temperature detector output signal failing HIGHz (1.0)
- a. An open circuit for an RTD.
- b. A short circuit for an RTD.
- c. An open circuit f or a thermocouple.
- d. A short circuit for a thermocouple.
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4.__BEB9I98_EBINglELgg_J23) __THEBMggYN8MIgg Page 24 JZZ1_SUD_G90E9 BEN 1g_J19*/.) (FUNpAMENTALg_gXAM1 OUESTION 4.21 (2.00)
STATE the type of protection provided by the following generator trip relays. (2.0) e, Negative Phase Sequence Relay. .
- b. Phase Angle Dif f erential Current Relay.
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- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 26 !
(33%1 QUESTION 5.01 (2.00)
' Northern States Power LER 88-001 was submitted to the NRC as a result of an investigation that determined inadequate net positive suction head (NPSH) could occur in the low pressure safetv injection system.
- a. Under WHAT plant conditions was it determined that inadequate NPSB could exist? (Include specific system lineup in your answer). (1.0)
- b. State how the resultant procedural changes affect operation of the Plant. (1.0) i i
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- 5. Page 27 l EMERGENCY AND ABNORM &L PLANT EVOLUTIONS
( 33%)
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1 QUESTION 5.02 (1.50)
You are performing ES-0.2, NATURAL CIRCULATION COOLDOWN, following a loss of offsite power, when you notice that actual pressurizer level is ;
varying by large amounts, inconsistent with your cooldown rate. j i
- n. What is the problem? (0.75) !
- b. What action would you take to regain normal pressurizer 3evel indication, assuming no situation existed that would require immediate cooldown and depressurization? (0.75) j
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- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 28 -
133%)
1 QUESTION 5.03 (1.00) .
Which ONE of the followints statements is correct:
Tha Technical Specification bases for restricting the quantity of j radioactive material contained in condensate storate tanks-to less than or equal to 10 curies provides assurance that ..: (1.0)
- a. ...the dose received by an individual standing beside the tank will j not exceed the 10CFR20 limit of 1 mR/hr. j d
- b. ...a release of the tank contents to atmosphere will not exceed the ALARA goals for gaseous waste release.
- c. ...a release of the tank contents to atmosphere will not exceed the l 10CFR20 limits for a release to an unrestricted area.
- d. ...t.he dose received by an individual standing beside the tank will not exceed the ALARA goal of 1 mR/hr.
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Page.29 j
- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS (33%1 l
I QUESTION 5.04. (1.00)
Which ONE of the following statements is correct:
During a natural circulation cooldown, the steam generator levels are li enintained at the no-load value...: (1.0)
- a. ...to minimize the effects of a LOCA should one occur.
- b. ...to provide a stable heat sink for decay heat removal.
- c. ...to conserve inventory in the condensate storage tanks during the cooldown.
- d. ...to maintain the cooldown rate at less than 25 degrees F per hour in the RCS cold legs.
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- 1?I31 DUESTION 5.05 (2.00)
No. During refueling operations, which DNE-of the f ollowing accurately '
represents the Technical Specification Residual' Heat Removal System.
(RHRS) operati onal requirements. (1.0)-
While f uel i s bei ng ' moved. . . :
- 1. . . .TWD RHR pumps must be in operation.
- 3. ...ONE RHR pump must be operable'and in operation,-but:may be taken out of service to f acilitate movement of .f uel.
Of.uP 4 .o m hoo(-
- 4. ...the RHR pump in operation may.be taken out of service no:
more than one hour'per day,
- b. If the Technical Specification RHRS operational requirements can'not-be maintained, which ONE of .the f ollowing actions must. be taken:. (1.0)
- 1. Suspend all fuel handling operations in containment.
- 2. Obtain. ref ueling containment integrity fwithin .1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.-
- 3. Increase baron concentration to greater than 2000 ppm.
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- 4. Generate an alternate cooling supply within one hour..
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Page 31
- h. EtfERGENCY AND_6BNORMAL PLANT EVOLUTIONS 133%)
1 QUESTION 5.06 (1.00) Te&$(Ol L[\C(%h'O()
Which ONE of the following comprises the\ requirements for the Nuclear Flux j
Source Range Instrumentation: (1.0) i i
- n. Source range instrumentation cannot be bypassed until 2 of 2 !
intermediate range channels read greater than 10 (-10) amps.
- b. When the minimum number of operable source range' channels cannot be satisfied, the RCS average temperature must be maintained below ,
S47 degrees F.
- c. An overlap of one source range deiw.: tor and one intermediate range detector will permit startup of tira reactor.
- d. During fuel movement, one source range channel must be operating at ,
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' 5 ~. EMERGENCY AND ABNORML PLANT EVOLUTIQ M 12311 QUESTION 5.07 (2.00)
- a. For Unit 1 PI, which ONE of the following statements describes; ;j why a steam flow / feed flow mismatch may be observed in a steam generator with a constant 44% level and a tube rupture prior to ,
a reactor trip: (1.0)
- 1. The level control' system is flow dominant.
- 2. The level control system is level dominant.
- 3. There is more water entering the steam generator than exiting as steam.
- 4. Steam flow is equivalent to rupture flow plus feedwater flow.
- b. Which ONE of the following identifies the parameters that determine the equilibrium RCS pressure following a steam. generator tube rupture and reactor trip. (1.0)
- 1. a. size of tube failures b, capacity of SI system
- c. RCS cooldown rate
- 2. a. size of tube failures
- b. number of reactor coolant pumps in operation
- c. pressure in intact steam generator
- 3. a. RCS cooldown rate
- b. number of auxiliary feedwater pumps in operation
- c. capacity of SI system
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- 4. a. number of reactor coolant pumps in operation
- b. number of auxiliary feedwater pumps in operation
- c. pressure in intact steam generator 1
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- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS (33%)
)
QUESTION 5.08 (1.00) fou are given the following summary of Critical Safety Function Status Trees while performing IE-2, " Faulted Steam Generator Isolation". If ALL procedures were to be used, which ONE of the following represents the order in which they would be parformed: (1.0) ;
- 1. Heat Sink Red FR-H.1
- 2. Core Cooling Yellow FR-C.3
- 3. Suberiticality Orange FR-S.1
- 4. Inventory Yellow FR-I.3
- 5. Containment Red FR-Z.1
- 6. Integrity Orange FR-P.1 .l 1
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- n. 1,5,3,6,2,4
- b. 5,1,3,6,4,2 i
- c. 1,5,6,3,2,4
- d. 5,1,6,3,4,2
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- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS 11111 QUESTION 5.09 (1.00)
Which ONE of the following indications would verify natural circulation flow in procedure 1ES-0.1, Reactor Trip Recovery: (1.0)
- n. 1. S/G pressures - stable or decreasing
- 2. RCS pressure - stable or increasing
- b. 1. S/G pressures - stable or increasing
- 2. RCS subcooling (based on core exit T/C's) - > 20 degrees F
- c. 1. S/G pressures - stable or decreasing
- 2. RCS pressure - stable or decreasing
- d. 1. S/G ressures - stable or decreasing
- 2. RCS subcooling (based on core exit T/C's) - > 20 degrecs F
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Es__ECEEEENGl_6dE_6BM9806b_Ek6Ml_EM96M11gyg 112EL i
DUESTION 5.10 (2. 00)-
Using the attached EPIP, F3-2, classif y the f ollowing TWO events. Consider-each (a & b) separately. Give the minimum classification which MUST be declared (0.5 each) and reference the initiating condition number'froniF3-2 .j (0.5 each).
At 0826: a reactor. trip occurs l a.
At 0828: the High Pressure Injection System automatically actuates .
and injects into the RCS At 0845: Pressurizer level is 0% l RCS pressure is 1500 psig Containment Pressure is 6-psig and increasing :
At 0915: Pressurizer level is 0%
RCS pressure is 1200 psig Containment pressure is B psig and increasing Subcooling margin is O degrees F.
- b. A " seismic event" is detected on the station seismic instrumentation.
A fuel assembly being removed f rom the core inside containment is dropped, resulting in damagn to the assembly. ;
R7 increases to 400 mR/hr.
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- LLDEBGENCY AND ABNORMAL' PLANT EVOLUT.lQNS Page 36
~C4011 QUESTION 5.11 (2.00)
.a. What are the TWO transition / entry conditions for EOP 1FR-S.1, Response To Nuclear Power Generation /ATWS? (1.0)
- b. While responding to an ATWS, the reactor and turbine will not trip manually or automatically. What is the basis for not mandally initiating Safety Injection under these conditions? (1'.0)
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- 5. EMEEGENCY AND ABNOB!iAL PLANT EVOLUTIONf2 Page 37 ,
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QUESTION 5.12 (1.50)
Fill in the blanks: (Each blank may require more than one word) (1.5)
The limitations on the specific activity of the RCS ensures that the resulting (time) doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a cccident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of .
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- 5. EMERGENCY AND ABNOBti&L PLANT EVOLUTIQHS Page 38 133%)
QUESTION 5.13 (2.00)
- c. An RCCA has just dropped into the core while at 100% power. Based on the accident analysis, what Reactor Protection feature will trip !
the reactor if a trip occurs? .(1.0) a
- b. If a reactor trip DID NOT occur, and the rod will not be. retrieved immedir.t.cly, which ONE of the following includes actions that must be taken oy the operator in accordnace with Operating Procedure CS, Control Rod System: (1.0)
- 1. a. Set NIS power range overpower trip high range to 60% rated power.
- b. Reduce power to 50%.
- 2. a. Reduce reactor power to 80%.
- b. Maintain delta I within target band.
- 3. a. Maintain delta I within target band.
- b. Reduce power to 50%.
- 4. a. Reduce power to 80%
- b. Set NIS power range overpower trip high .to 90% rated power.
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UUESTION 5.14 (1.50) 4 I'
Given the f ollowing RCCA withdrawal events according to the accident anal ysi s , what reactor protection system f eature is assumed to trip the reactor?
- a. Uncontrolled RCCA withdrawal from a ze.o hot power condition.. (0,5) !
FAST 400 pcm/sec) uncontrolled RCCA withdrawal at 100% power. (0.5) b.
I' SLOW (10 pcm/sec) uncontrolled RCCA withdrawal at 100% power. (0.5) c.
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EMERGENCY AND_@pNORMAL_ PLANT _ EVOLUTIONS Page 40 S.
12231 DUESTION 5.15 (1.00)
Following a Main Steam Line Breair, it must be determined if SI should be stopped. Which DNE of the following plant status cituations would allow the oper ator to terminate SI: (1.0)
- a. Containment Pressure = 6 psig RCS Subcooling based on T/C's = 50 degrees F AFW f l ow to intact S/G = 300 gpm Narrow range level in intact S/G = 12%
Pressurizer level a 10%
RCS pressure is 2200 psig and increasing
- b. Containment Pressure = 8 psig RCS Subcooling based on T/C's = 100 degrees F AFW flow to intact S/G = 100 gpm Narrow range level in intact S/G = 5%
Pressurizer level = S%
RCS pressure is 1900 psig and increasing
- c. Containment Pressure = 2 psig RCS Subcooling based on T/C's = 40 degrees F AFW flow to intact S/G = 150 gpm Narrow range level in intact S/G = 15%
Pressurizer level = 3%
RCS pressure is 1900 psig and stable
- d. Containment Pr essure = 4 psig RCS Subcooling based on T/C's = 30 degrees F AFW flow to intact S/G = 50 gpm Narrow range level in intact S/G = 20%
Pressurizer level = 10%
RCS pressure is 2200 psig and stable i
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OUESTION 5.16 (2.00)
- c. IDENTIFY the. Emergency Operating Procedures that would be used during
'the RECOVERY and COOLDOWN from a loss of all offsite power. (Assume .
i no further complications occur). ( 1. 0 )
- b. Five minutes into a loss of offsite power event, the shift-supervisor recognizes that charging is in service, but letdown 'is isolated, even l though pressurizer l evel remained-above the letdown isolation setpoint throughout the transient. WHAT .specifically caused letdown j to isolate? (Assume normal eler.trical lineup prior to initiation of event.) (1.0) l i
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~ 22__EUEB9EUGY_6HD_eEN9BDBL_EkBNI_EV9kUTIONS d22Z4 OUESTION' 5.17 ( 1. 00) '
- Which DNE. of f the f ollowing conditions, combined with a Hi Steam Flow in-DNE Steam Generator,~would result in the automatic
- closure of.only ONE MSIV: .(1.0)
, m. Saf ety Injection ' Signal
.b. Hi-Hi Containment Pressure and Safety Injection Signal
- c. Lo-Lo Tavg and Safety Injectron Signal
- d. La Tavg and Hi Containment Pressure i
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- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 43 (33%)
QUESTION 5.18 (1.00)
Which ONE of the following statements is correct concerning the design of ths Auxiliary Feedwater (AFW) System: (1.0)
The AFW System is designed. . . :
- a. ...to automatically actuate on a loss of one main feedwater (MFW) pump to minimize the affects of the reduced MFW flow.
- b. ...to remove decay heat at a rate equivalent to 10% full power.
- c. ...to prevent the pressurizer safety valves from lifting following a blackout.
- d. ...to return steam generator level between 10% and 50% within 10 minutes following a reactor trip.
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- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 44 (33%)
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i QUESTION 5.19 (2.00) 1 Tha following questions concern the plant design features to mitigate the H environmental consequences of a LOCA. ,
- n. Disregarding fuel cladding and RCS piping / system, what art the TWO !
barriers in series with fission product leakage that minimize leakage to the environment? (1.0)
- b. Which ONE of the following includes the functions of the annulus in mitigation of a LOCA: (1.0)
- 1. a. Bolds leakage for decay. '
- b. Dilutes leakage prior to release.
- 2. a. Holds leakage for decay.
- b. Reduces pressure in containment. ;
- 3. a. Dilutes leakage prior to release. l
- b. Minimizes the radiation effects of electrical equipment in '
the annulus.
- 4. a. Reduces pressure in containment.
- b. Minimizes the radiation effects of electrical equipment in the annulus.
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.Page 45-5 .__Et!EEEENGY_6ND_6EN9BdBL_EL ANT __ EVQLUT1995 '
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QUESTION 5.20 (1.50) j An RCS. leak is in progress while at 50% power.
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- a. At what point would the RCS leak be. classified as a:LOCA? .(0.50) . j 4
- b. If 'not manually tripped, which DNE of the following automatic reactor j trip signals would trip the reactors' ( 1. 0) - i d
1.- Loss of reactor coolant flow. f 1
- 2. Low pressurizer pressure.
- 3. Low pressurizer level.
- 4. Overpower delta T.
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Page'46 5 __ENEBGgygy_8Mp_8BygSUS6_E66MI_EVg6gIjgyS (33%)
(
l DUESTION 5.21 (3.00) i Match ALL of the f ollowing initial and subsequent actions to be taken. in l COLUMN B to the associated Radiation Monitors in COLUMN A, should they l alarm. COLUMN B actions can be used more than once or not at all.
COLUM A may have more than one answer. (3.0) {
l COLUMN A COLUMN B
- a. 2-R-9, Reactor 1. Close containment purge valves.
Coolant Letdown Line Area Monitor. 2.- If radi ati on level is 15 R/hr, Unit 2 shall be placed in cold shutdown.
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- b. 2-R-11, Containment / 3. Turn on all turbine b1dg roof Shield Bldg Vent Air ventilators.
Particle Monitor.
- 4. Sample mixed bed demineralized f or both influent and effluent.
- c. 2-R-12, Containment /
Shield Bldg Vent 5. Check steam lir e radiation monitors.
Radioactive Gas Monitor. _____
- 6. Survey auxiliary building.
- d. 2-R-15, Condenser Air 7. Stop all releases in progress.
Ejector Radioactive Gas Monitor. _____
- e. 2-R-19, Steam Generator Blowdown Liquid Monitor.
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, .I 62.__EL6MI_EYSIEME_19931_6MD_ELONI WIDE __QgNERig Page 48 BE@EQNEl@l(ITIE@_11l i DUESTION 6.01 (1.00) Which DNE of the following will result in a positive reactivity addition J to the Reactor Coolant System: ( 1'. 0 ) J
- a. Decrease in main feedwater. temperature.
b.. Decrease in main f eedwater flow. I
- c. Increase in reactor cool ant system temper ature.
- d. Decrease in main steam flow.
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- 6. Page 49 PL6NT SYSI M _[30%1_AND PhANT-WIDE GENERIC RESPONSIBILITIES (13%)
QUESTION 6.02 (1.00) Which ONE of the following completes this statement: (1.0)
"During a plant cooldown (RCS temperature is 325 degrees F), flow from the Residual Heat Removal System (RHRS) to the. Reactor Coolant System is designed to be ...:
- o. ... constant (around 2000 gpm) AND this total flow is maintained by automatically controlling the RBRS heat exchanger flow bypass valve (CV-31237)."
- b. ... constant (around 2000 gpm) AND this flow is maintained by automatically controlling the RHRS heat exchanger outlet valves (CV-31235 and CV-31236)."
- c. ... varied (to a flow that meets the desired cooldown rate) AND this flow is maintained by controlling the RHRS heat exchanger flow bypass valve (CV-31237)."
- d. ... varied (to a flow that meets the desired cooldown rate) AND this flow is maintained by controlling the RHRS heat exchanger outlet valves (CV-31235 and CV-31236)."
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- 6. PLANT' SYSTEMS L30%)__AND PLANT-WIDE _ GENERIC Page 50 RESPONSIBILITIES (13%1 QUESTION' 6.03 (2.00)
Concerning the rod position indication system, which ONE of the following (in a. and b.) correctly describes the system *
- a. The Baux Demand Position Indication System (Step Counters)... (1.0) !
- 1. ...is'not very accurate, but very reliable. I
- 2. ...is accurate to within +/- 1/2 step. j
- 3. ...provides indication of actual rod group height'b7 Oounting pulses.
- 4. ...is unable to detect a stuck rod. i l
- b. The Individual Rod Position Indication System... (1.0) ,
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- 1. ...is accurate to within +/- 1% of full rod travel. l
- 2. ...is very accurate and very reliable.
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- 3. ...contains primary coils in series.
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- 4. ...contains one position detector for each group of rods.
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- 6. PLANT SYSTEtjS_{30%1 AND PLANT-WIDE GENERIC Page 51 RE_SEONSIBLLITIES (13%1 QUESTION 6.04 (2.00) i Unit 1 is in cold shutdown. An operability test of the No. 12 Motor Driven Auxiliary Feedwater Pump (AFWP) has just been performed, and the control switch placed in " Shutdown Auto", in accordance with the procedure. While the #12 St.eam Generator is being drained to allow tube plugging, the #12 AFWP automatically starts,
- n. What automatic start signal caused the pump to start? (1.0)
- b. Procedurally, what action (s) must the operator take to prevent recurrence of this event? (1.0).
.. - __ __ _ _ ___ _f _.
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Page 52 6z__ELEMI_E1SIEMS_1ggKl_6MQ_E66BI BlqE_GEggBig ; BEEEOUSIB16111ES_11s31 -i l l i DUESTION 6.05 (2.50)
- a. What are TWO Engineered Saf eguards System initiation signals that (1.5) mitigate pressurized thermal shock to the reactor vessel? {
- b. Which DNE of'the following sets of RCS' conditions would result in a challenge to the Integrity Critical Saf ety Function: (1.0)
- 1. Rapid RCS cooldown OR Rapid RCS' pressure decrease while the j RCS is hot.
- 2. Rapid RCS heatup OR Rapid RCS pressure increase while the RCS is cold-.
- 3. Rapid RCS cooldown OR Rapid RCS pressure increase while the ,
RCS is cold. .)
- 4. Rapid RCS heatup OR Rapid RCS pressure decrease while the RCS is hot.
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- 6. PLANT SYSTEMS (30%) AND PLANT-W1DE GENERIC' Page 53 -
RESPONSIBILITIES (13%1 I 1 I QUESTION 6.06 (1.00) ! Which ONE of the following statements is correct concerning the Emergency Diesel Generators: If a complete loss of DC power occurred, and the air start solenoids are open....: (1.0) I a. ...the diesel generator would require manual flashing using the field l flash pushbutton located on the Engine Control Panel,
- b. ...a 12 volt car battery connected to terminal board TS4 would supply enough current to flash the field.
- c. ...the generator would produce voltage after starting locally if enough residual magnetism were present.
- d. . . .the generator controls are prepared for manual start by setting the governor synchronizer control to " Maximum". 1 1
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L, PLANT SYSTEMS _[30%) 6ND PLANT-WIDE GENEBLQ Page 54 RESPONSIBILITIES (123il QUESTION 6.07 (1.00) Which ONE of the following statements is CORRECT regarding the Technical Specification pertaining to the Auxiliary Feedwater System: (1.0)
- a. The condensate storage tank supply cross connect valves to the auxiliary feedwater pumps must be tagged open.
- b. For single unit operation, the turbine-driven pump and motor-driven pump associated with the unit must be operable.
- c. A minimum of 100,000 gallons of water in the condensate storage tanks is specified to ensure sufficient decay heat removal for one reactor for 36 hours after shutdown.
- d. For two-unit operation, both turbine-drivsn pumps and one motor-driven pump must be operable.
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'6 __ELONI_EYEIEUE_129%l_0ND_EkeMI: WIDE _EENESIC EEEEONElalbillES (1731
'OUESTION 6.08 (1.00)
Which ONE of the f ollowing statements is CORRECT with ' regard to the Main Fordwater (MFW)~ Pump control: (1.0)
- a. P>ior to' starting the first-MFW pump, two out of three condensate pumps suust be in operation, providing a MFW pump suction' pressure greater than 220 psig.
- b. Lube oil pressure less than 6 psig on the #11 MFW pump _will generate an automatic trip of the pump.
- c. The seal water supply pressure to the #21 MFW pump must be greater than 220 psig bef ore : the pump can be manually started.
d.. A high level en the #11 steam generator instrumentation (2/3 at=50%) will generate an automatic _ trip of the #11 MFW pump. l 1
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- b. _E60BI_SYSIEUS 12931-@Np_Eh6NI:WIDg_ggggBIC BESEONSIE16111EE_11s31 !
DUESTION 6.09 (1.00) Which ONE of the f ollowing conditions would result in the actuation of the Containment Spray System Pumps: (1.0) )
}
- c. Containment pressure is at 17 psi g , the " Local-Remote" selector ] '
switches are in the " Remote" position, and the operator depresses the " start" pushbuttons by the pumps. l I
- b. One out of two (1/2) Containment High-High Pressure signals at 23 psig ]
l on two out of three (2/3) sets of containment pressure detectors. ! l
- c. The " Local-Remote" csitch is in the " Local" position and two out of two (2/2) Control Board manual " Containment Spray Actuation" switches are placed in " ACTUATE".
- d. The containment spray "Stop-Normal-Start" control switch is in the
" NORMAL" position, and a "P" signal is received.
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- 6. P's&NT_ SYSTEMS _L30%) AND PLANT-WIDE GENERIC Page 57 RESPONSIBILITIES (13%1 i l
QUESTION 6.10 (1.00) Which ONE of the following statements is MOST accurate concerning the transfer of the ECCS System to the recirculation mode:, (1.0)
- n. The containment spray pump is placed in the pull-out position after l the containment spray signal is reset to prevent cycling of the pump i discharge va?ve when the pump is shutdown.
- b. While making the transfer, complete valve travel for the SI test line to the RWST valves must be verified prior to proceding with the transfer to prevent contamination of the RWST.
- c. The SI test line isolation valves are closed to prevent a release of radioactive gases to the Auxiliary Building environment.
- d. The RHR pump suction valves from Sump A are opened to suppl.v suction to the RHR pumps.
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- 6. PLANT SYSTE M _QO M _AND PLANT-WIDE GENERIC Page 68 RESPONSIBILITIES (13%)
QUESTION 6.11 (1.00) Which ONE of the following is the MOST accurate with regard to the Incore Instrumentation System: (1.0) ,
- o. The incore thermocouple are located over fuel assemblies in the core to measure fuel assembly outlet temperatures.
- b. The incore thermocouple are inserted into the reactor core through tub s and measure the coolant flow temperature axially through the core.
- c. The incore thermocouple are located in the center of the core to measure the maximum fuel assembly temperatures,
- d. The incore thermocouple are located beneath the ft- assemblies to I
measure the coolant flow temperature after it passes through the core. 1 (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
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- 6. PLANT SYSTEtiS__L30_%1_AND PLANT-WIDE GENERIC Page 59.
BESPONSIBRITIES (13%) QUESTION 6.12 (1.00) Which ONE of the following emergency diesel protective trip signals is i cetivo during a safety injection: (1.0) ]
- a. Crank case high pressure !
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- b. Reverse current
- c. Differential current
- d. Low lube oil pressure .) l I
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Page 60 6 __ELBMI_SYE1 EMS (gp31_@NQ_E(@NI;Wlpg_GENEBIC BEEE9NglBibillg@_ll}Z1 I OUE3'l ION 6.13 (2.00) ]a Concerning the Main Steam System, match the f ollowing components in' , COLUMN B with the associated descriptions in COLUMN A. COLUMN B COLUMN A may.have components may be used more than once-or not at all.
. (2. 0) more than one answer. - \
COLUMN A COLUMN B j
- a. Used for RCS decay heat removal 1. ' steam flow nozzle with a main steam line rupture downstream of the MSIV. _____
- 2. S/G PORV's }
- b. Capable of stopping reverse steam 5. safety valves flow. _____
- 4. MSIV's a l
- c. Located inside the reactor building. _____ 5. check valve -immediately i' downstream of MSIV.
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- 6. PLANT SYSTEMS (30%) AND_ PLANT-WIDE GENERIC Page 61 EEEPONSIBILITIES (13%1 I
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-QUESTION 6.14 (1.00) ;
.i Which ONE of the following statements correctly describes the Technical i j
Specification basis for the.following: During refueling operations, the reactor cavity is filled with borated , water to a minimum depth of 23 feet to...: (1.0) !
- n. ... ensure radiation at the surface remains below 50 mR/hr during fuel transfer. j 1
- b. ... ensure the assemblies remain covered during the transfer.
- c. ... ensure the reactor won't go critical.
- d. ... ensure proper operation of the cavity leakage detection system.
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- 6. PLANT SYSTEMS (30%) - AND' PLANT-WIDE GENERIC Page 62 !
RESPONSIBILITIES (13%1 { I i < QUESTION 6.15 (1.00) Which ONE of the following statements is correct concerning the Emergency Dional Generator Fuel Supply: (1.0)
- c. A fuel supply of 70,000 gallons is available in the interconnected storage tanks to supply one train of diesel . generators for' 14 days.
- b. A fuel supply of 70,000 gallons is available in the interconnected storage tanks to supply one train of diesel generators for 7 days. )
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- c. A fuel supply of 35,000 gallons is available in the interconnected I ctorage tanks to supply one train of diesel generators for 14 days. ,
- d. A fuel supply of 35,000 gallons is available in the interconnected storage tanks to supply one train of diesel generators for 7 days. j 1
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Page 63 6 t__E66NI_SYgIgd@_1}QKl_6NQ_ELGNTrWlDg_QgNg31g EEEE9BelE16111ED_11231 l DUESTION 6.16 (1.00) Which ONE of the f c11owing statements is correct concerning Containment Integrity requirements during refueling operations: 01.0)
- a. With only new fuel in the reactor, containment integrity 15 required if Keff is 0.98. ;
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- b. Wi th old f uel in the reactor, containmentLintegrity.IG eequired if Keff is 0.93.
- c. With the reactor half loaded with old f uel, containment integrity IS NOT-required if RCS temperature remains below-120 degrees F.
- d. When the reactor is suberitical by at least 10% delta k/k, containment j integrity 19 NOT required, provided only one rod control c3uster assembly is manipulated at a time. l l
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l f, PLANT SYSTEMS (30%) AND PLANT-WIDE GENEB19 Page 64 RESPONSIBILITIES (33%1 i l QUESTION 6.17 (2.00) Concerning the Fire Protection System Jockey Pump: >
- c. WHAT signal results in the automatic start of the Jockey Pump? (1-0)
- b. During a loss of all AC power, what component provides water.to the fire protection system? (1.0) i 1
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- 6. PLANT SYSTEMG_[30%) AND_ PLANT-WIDE GENERIC Page 65 BEQP_QNS_lBILITIES (13%)
QUESTION 6.18 (1.00) An RCS heatup is commenced using the pressurizer heaters. Which ONE of. the following statements is correct regarding the heatup: (1.0)
- o. The RCS and pressurizer shall meet the water chemistry specifications for oxygen, chloride, fluoride, and lithium before the RCS temperature exceeds 150 degrees F,
- b. RCS pressure is controlled by adjusting the letdown pressure control valve.
- c. The pressurizer and RCS heatup rates shall not exceed 100 degrees F per hour,
- d. When the pressurizer temperature reaches saturation, the temperature downstream of the PORV will increase to the saturation temperature.
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- 6. EMN_I_SYSIEt1S (10%) AND PLANT-WIDE GENERIC Page 66 l' EESPONSIBlLITIES_llM1 4
i QUESTION 6.19 (1.00) ] Which ONE of the following statements describes the PRIMARY function of f.ha Pressurizer Level Control System: 1 Tha Pressurizer Level Control System maintains pressurizer level during 1 pow 3r operation to ...: (1.0) i
- c. ... maintain nearly constant RCS water mass. .
- b. .... prevent damage to pressurizer heaters by operating them uncovered,
- c. ... protect the pressurizer relief valves. ,
- d. ... prevent an excessive pressure decrecse on a large insurge of water.
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1 1 Page 67 6t__PL.6BI_EYSIEUE_10931_6SD_ELANT-Wlps_QEGE61Q
.BF3EQUE1E161ILEE_11231 DUESTION 6.20 (1.00)
Which DNE of the f ollowing statements is correct concerning the ! Reactor Protection System: (1.0)
- a. A low pressurizer pressure reactor trip will occur at 1900 ps.~g, with a .2/3 logic, to protect the core from DNB..
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- b. An undervoltage condition of 75% rated voltage on Bus 11, {
with a 1/2 per bus logic, will trip the reactor in anticipation l j of'a loss of coolant flow. -
- c. A low-low steam generator level' reactor trip will occur at 15% level, with a 2/3 logic, to protect the core from a loss of heat sink. .
- d. A high pressurizer level reactor trip will occur at 90% level, with i a 2/4 logic, to prevent the pressurizer safety and relief valves f rom lif ting.
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PLANT __ SYSTEMS _L303)_AND PLANT-WID_E GENERIC Page 68-
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-QUESTION 6.21 ,(1.00) i Which ONE of the following statements is correct.regarding the Reactor Vassel Level Instrumentation System (RVLIS): (1.0)
- n. Reactor. Vessel Dynamic Head indication is the actual diffe,rential pressure across the reactor core and internals.couspared to the anticipated differential pressure for. normal, single phase coolant I flow.
- b. Reuctor Vessel Upper Range indication displays the level in the reactor vessel above the loop cold leg connections. '
- c. During full flow conditions, RVLIS full range indications are inaccurate due to the lack of temperature compensation-with the reactor coolant pumps in operation,
- d. RCS Narrow Range Temperature is used as an laput to RVLIS with the reactor coolant pumps (RCP's) in operation'and RCS Wide Range Temperature is used as an input when the RCP's are not in operation. :!
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-l 6 PLANT SYSTEMS-(30%) AND_ PLANT-WIDE GENERIC Page 69 l
BESPONSIHIL}TJ_ES E (13%1 i i QUESTION 6.22 (1.00) j Which CNE of the following group of events comprises the design basis for the Emergency Core Cooling System: (1.0)
- c. LOCA, Loss of secondary coolant, Loss of both reactor coolant pumps, l and Loss of offsite power. j l
- b. LOCA, Steam generator tube rupture, Loss of offsite power, and Rod' ejection accident.
LOCA, Loss of secondary coolant, Steam generator tube rupture, and 1 c. Rod ejection accident.
- d. LOCA, Loss of secondary coolant, Rod ejection accident, and Loss of. .j both reactor coolant pumps. j
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i Pege"70 6 '. PLANT-SYSTEMS _L30%) AND PLANT-WIDE GENERIC LtESPONSIBILLTigS (13x> .
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1 QUESTION 6.23 (3.00) l ! Concerning the Steam Dump System,. HATCH the correct value from COLUMN B to' .I the statement in COLUMN A. COLUMN B values may be used more than once, or not at all. (3.0) COLUMN A COLUMN B -l
. o.- The steam dump system is initiated only when '1. 2.b%
a step load decrease exceeding or an equivalent ramp load decrease occurs. '2. 7.5%
- b. Two atmospheric dump valves are sized to 3. 10.0%
pass of full load' steam flow,
- 4. '15'0%1
- c. The condenser dump ~ valve in automatic is.
nized to pass of full load steam flow. 5. 20.04 6.' 30.0% (***** CATEGORY' 6 CONTINUED ON NEXT.PAGE:*****)
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- - - AND_PLGNT-Wigg_GENEBlQ Page 71 ),
BEgBON@l@l(lllES_11231 l l 4 l QUESTION 6.24 (1.00) Which DNE of the following statements is correct concerning' Independent 1 Verification perf ormed by- electricians: (1.0)
- c. The Shif t Supervisor is responsible f or ensuring that Ir: dependent j Verification is perf ormed by a competent electrician.
- b. If an Independent Verification would result in an exposure greater j than 10 mrem, physical Independent Verification is not required.
- c. The Independent Verification can not be performed by the first line supervisor. ;
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- d. Independent Verification need only be perf ormed on equipment that has~ .)
j been tagged out of service.
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L _EL6HT SYSTEMS _L' 1 0%) AND PLANT-WIDE GENEBLC Page 72. BESPONSIBiblIIES (13%1 QUESTION- 6.25 (2.50)
- a. The TWO main sources of radiation to personnel at Prairie Island are !
products and _ products. (1.0) --
- b. Which of the following'are ways the plant is physically operated to sinimize doses to personnel, as discussed in the ALARA program (there may be more than one answer): (1.5) !
- 1. Perform an intentional crud burst each outage. !
- 2. Continuous use of the cation bed domineralizer while at 100%
].j power. j
- 3. Minimize the number of cycles in which a fuel assembly is used.
- 4. Minimize hydrogen levels in the RCS. -
- 5. Minimize power rate changes.
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- 6. Minimize oxygen levels in the RCS. )
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DUESTION 6.26 (3.00) ; Concerning the Fire Protection Program, match ALL'the types of fire
-extinguishers in COLUMN B that can bs' used to extinguish the class of fire in COLUMN A. The extinguishers in CDLUMN B ray be used more than once or j COLUMN A may have more than one answer. (3.0) not at all. l COLUMN A COLUMN B
- a. Class "A" Fire 1. Halon Extinguisher )
Pressurized Water Extinguisher
- 6. Class "B" Fire _____ 2.
- c. Class "C" Fire _____ 3. Carbon. Dioxide Extinguisher d., Class "D" Fire _ 4.' Dry Chemical Extinguisher.
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H. PLANI _ SYSTEMS (30%) . AND PLANT-WIDE GENERIC Page 74 BESEQNSIBII.ITIES (13%1 , J l l QUESTION 6.27 (2.00) M2tch the following descriptions of areas in the plant in COLUMN B with the , designated area in COLUMN A. The descriptions in COLUMN B may be used mo.e i than once or not at all. (2.0) ] COLUMN A COLUMN B
- o. Radiation Area 1. Area with a continuous dose rate of 3.5 mR/hr.
- b. High Radiation Area
- 2. Area where a frisker monitors
- c. Very High Radiation Area 50-100 cym.
- d. Contaminated Area 3. Barricaded or roped off area. 'i 1
- 4. Airborne Radioactivity it.
concentration in excess.of 25% of the limits for restricted' areas as established in 10CFR20
- 5. Spent resin tank room.
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- 6. PLAHI SYSTEMS (301r,) AND PLANT-WIDE GENERIC Page 75 BESPONSIBILIIIES (130 QUESTION 6.28 (2.00) cQUE; TION Stoto the TWO situations that allow the temporary removal of HOLD and SECURE cards and temporary restoration of equipment.
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- 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIG Page 76 RESPONSIBILITIES (13%1 1
QUESTION 6.29 (2.00)
- c. What is the basis for adding chlorine to the Cooling Water' System? (0.75) b.. What is the greatest health hazard of chlorine? (0.75)
- c. Should chlorine contaminate the skin or clothing, what actions should be taken? (0.5) l 2
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- 4. REACTOR _ PRINCIPLES (7%1 THEM QDYNAMICS Page 77 .
17%) AND COMPONENTS __{10%) 1 FUNDAMENTALS EXAM 1 l i i I i ANSWER 4.01 (1.00) J CANSWER
- a. Critical rod height does NOT depend on how fast control rods are witbdrawn. (1.0)
REFERENCE
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CREFERENCE ] Operating Procedure Appendix CIA, Reactivity Calculations ! 192008K101 192008K105 ..(KA's) ANSWER 4.02 (2.00) CANSWER P = P(o)10[to the exponent of SUR times TIME (in minutes)] SUR = log P/P(o) divided by TIME j i SUR = los 5000/1000 (divided by .5 min) l SUR = log 5 (divided by .5 min) SUR = .7 (divided by .5 min) = 1.4 DPM (1.0 pt for using correct equation, 1.0 pt for correct answer) REFERENCE CREFERENCE Westinghouse Fundamentals of Nuclear Rx Physics, pg. 7-19 thru Y-20. 192003K105 ..(KA's) ANSWER 4.03 (1.00) CANSWER
- c. Void, MTC, Doppler (1.0)
REFERENCE cREFERENCE Waptinghouse Reactor Core Control for Large PWR's, pg. 3-29 thru 3-41. 192004K108 ..(KA's) y g g[ ab:tv us U 2 .1 (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
- 4. REACTOB_ ERIN _glPLES (7%1_ THERMODYNAMICS Page 78 17%) AND COMPONENTS (10%) (FUNDAMENTALS _ EXAM)
ANSWER L. 04 (1.00) OANSWER
- b. When a reactor is shutdown after 120 days of operation at 100% power, zenon burnup effectively stops while the decay of iodine continues.
Therefore, the xenon concentration starts to increase. (1.0) 3EFERENCE OREFERENCE Westinghouse Rx Core Control For Large PWR's, pg. 4-20 thru 4-23. 192006K103 ..(KA's) ANSWER 4.05 (1.00) DANSWER
- d. Core age INCREASES (1.0)
REFERENCE J OREFERENCE Westinghouse Fundamentals of Nucle <r Rx Physics, pg. 6-20 thru 6-23. 192007K104 ..(KA's) j ANSWER 4.06 (1.00) l OANSWER
- d. The measurr 7f a reactor's departure from criticality. (1.0)
REFERENCE OREFERENCE I l Westinghouse Rx Core Control For Large PWR's, pg. 2-5 thru 2-6. 192002K111 ..(KA's) 1 (***** CATEGORY 4 CONF 1NUED ON NEXT PAGE *****)
l 4. REACIO_R__ PRINCIPLES (7%_,l_.TLIRtj0 DYNAMICS Page 79 l J7%)_AND COMPONENTS (10%) (FUND &JENTALS' EXAM 1
-ANSWER 4.07 (1.00)
OANSWER
- 1. a. Convection (0.34)
- 2. b. Radiation (0.33)
- 3. c. Conduction (0.33)
REFERENCE OREFERENT i Westinghouse Thermal-Hydraulic Principles &.Apps to the PWR I & II, pg. 3-101. 193007K101 193007K104 ..(KA's) ANSWER 4.08 (1.00) OANSWER i a. Hil-Ductility Temperature (accept NDT, RT-NDT, NDTT) (0.5) 1
- b. Cooldown (0.5)
REFERENCE
- REFERENCE Westinghouse Thermal-Hydraulic Principles & Apps To The PWR I & II, pg. 13-61 thru 13-62.
193010K102 193010K104 ..(KA's) l ANSWER 4.09 (1.00) 0 ANSWER
- c. a zircalloy-water reaction is accelerated at temperatures above 2200 degrees F. (1.0)
REFERENCE OREFERENCE Westinghouse Thermal-Hydraulic Principles & Apps To The PWR.I & II, pg. 13-16 thru 13-17. 193009K105 ..(KA's) (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
l
\
.- ]
- 4. REACTOR PRINCLELES (7%) THERMODYNAMLQS Page 80 j (7%1 AND COMPONENTS (10%) { FUNDAMENTAL.S EXAM) ]
l
.l ANSWER 4.10 (1.00)
CANSWER
- c. As pressurizer pressure increases, the DNBR increases. (1.0)
REFERENCE . CREFERENCE Westinghouse Thermal-Hydraulie Principles & Apps To The PWR I & II, j pg. 13-22 thru 13-24. j 193008K105 -,(KA's) ! l PNSWER 4.11 (1.00) CANSWER 0 ANSWER ;
- d. 1. Co1ntrol rods in a single bank move together with no individual I rod insertion differing by more than 15-inches from the bank demand position.
- 2. Control bank insertion limits are not violated. '
- 3. Axial power distribution control procedures.are observed.
(1,0) REFERENCE CREFERENCE '
- REFERENCE Toch Spec pg. 3.10-10.
193009K107 ..(KA's) ANSWER 4.12 (2.00)
- 0 ANSWER l a. Prevents hydrogen effervescence (0.25) on a rapid depressurization (0.25) which could blow water out of the reference leg (0.25) and give erroneous level indication (0.25).
- b. The difference in pressure (0.50) applied to the two bellows (0.50).
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
- 4. RE6CTOR PRINCIPLEG_Ll%1_THERMQDYN6MLCS Page 81 (7%) AND COMPONENTS (10%1_LFQ8[4RIENIALS EXAM 1 REFERENCE CREFERENCE Lesson Plan P8170L-006, pg. 10 of 43 and 12 of 43.
193001K103 ..(KA's) I' ANSWER 4.13 (1.00)
)
CANSWER l
- b. Circulating water outlet %<,aperature will decrease (1.0)
REFERENCE CREFERENCE Westinghouse Thermal-dydraulic Principles and Applications to the ; Pressurized Water Reactor II, 1982, pg. 9-19 l 191006K112 ..(KA's) ANSWER 4.14 (1.00) CANSWER
- d. Pressure increases, and the velocity decreases as the diameter of the venturi increases. (1.0)
REFERENCE
- REFERENCE Westinghouse Thermal-Hydraulic Principles and Applications to the Pressurized Water Reactor II, 1982, pg. 11-22.
191002K101 ..(KA's) ANSWER 4.15 (1.00)
~
cANSWER
- a. When a pump is started, the NPSH will decrease by the amount of the pressure drop in the suction piping. (1.0)
REFERENCE CRT.FERENCE Westinghouse Thermal-Hydraulic Principles and Applications to the Pracsurized Water Reactor II, 1982, pg. 10-55. 191004K101 191004K105 ..(KA's) [ (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
L RE6CIQB_EBIHgIELES_11%1_ILEBMODYN_AMICS Page 82 17_1_AND _% COMPONENTS (10%) (FUNDAMENTALS EXAM 1 l At:SWER 4.16 (1.00) l 1
.j 0 ANSWER
- d. One advantage of recirculation flow in a steam generator is that 1 as the level in the d ancomer increases, the delta P driving head ,
.)
increases. (1.0) REFERENCE
- REFERENCE Westinghouse Therme.-Hydraulic Principles and. Applications to the' q Pressuri_ed Water Reactor I, 1982, pg. 5-20 thru 5-25. i 191006K103 ..(KA's) i ANSWER 4.17 (1.00) ,
CANSWER
- d. The code safeties DO NOT have to be combined with the atmospheric dump valves to have a capacity of 300%, with relief settings from 1075 psig to 1129 psig. (1.0)
REFERENCE , CREFERENCE System Description B-27, Main and Auxiliary Steam System, pg. 7 of 18. 1910G ' F.101 ..(KA's) ANSWER 4.18 (1.GO) CANSWER DbD
. Indicated steam generator pressure will be E0m!St.han actual steam C, generator pressure.
REFERENCE l CREFERENCE l Westinghouse Thermal-Bydraulic Principles and Applications to the Fransurized Water Reactor II, 1982, pg. 11-11. 391002K111 ..(KA's) (***** CATEGORY 4 CONTINUED ON NEX1 PAGE *****)
ll a Page'B3 1 at_ _BE9910B_EBIGGIELES _- -( 7'/ l_IHERMQDYN AdigS , IZKl_6MR_G9dEONENIE_11931_1EUNp6 DENTAL @_g18d1
'l I
ANSWER 4.19 (1.00)
"q
- ANSWER j
- a. Dif ferential pressure across the flow venturi -in the steam line is .j propor ti onal to the volumetric flow rate. Volumetric flow rate (1.0) is compensated with the fluid density to provide mass. flow rate. a l
REFERENCE
]
OREFERENCE Lesson Pl an F8174L-OO6, Steam Generator Level Control System, pg. 20 of-52.- 191002K102 ..(KA's)
~
ANSWER 4.20 (1.00) 1 cANSWER
- o. An open circuit for an RTD. (1.0)
REF ERENCE OREFERENCE Westinghouse Thermal-Hydraulic Principles & Apps To The PWR I & II, _ l pg. 11-4. j 191002K114 ..(KA's) I
.i ANSWER 4.21 (2.00) ]
.1 l
OANSWER ) CANSWER i c. Protects rotor temperatures from increasing (0.5) to the point of heat damage (0,5). !
- b. (Protects the generator) against internal faults'(1.0).
REFERENCE oREFERENCE CREFERENCE System Description B-22B, Main Generator and Exciter, pg. 15 of 20 thru 16 of 20. 191005K111 ..(KA's) ) l l (***** END OF' CATEGORY 4 *****)
l 0 1 Page 84 ] Qu__gbgBgESCy_6SD_6pygBU66_PL6NT_EygLUTIgNS j 132E1 ANSWER 5.01 (2.00) cA,NSWER
- c. When the RHR pumps are in the recirc modeL(0.25). (f ollowing a large !
break LOCA), supplying low head upper plenum vessel injection (0.25), high head cold leg vessel injection (0.25), and containment spray (0.25) (si m_ ^ t aneousl y) . 1
. The simultaneous use of . low head upper plenum injection (0.5) and f b.
1 cold leg injection is not allowed (0.5). 1
)
REFERENCE 1 eREFERENCE Northern States Power LER 88-001 1 OOOO11K312 ..(KA's) 3 ANSWER 5.02 f.1. 50 )
~
eANSWER
- a. You have voiding (0.75) (in the reactor vessel or core) i
- b. Repressurize the RCS (to collapse potential voids) (0.75) ;
REFERENCE OREFERENCE EDP 1ES-0.3A, Natural Circulation Cooldown With CRDM Fans, pg. O ' or 15 I (Step 17). ' OOOO55K302 ..(KA's) I l l ANSWER 5.03 (1.00) , i l cANSWER 4
...a .elease of the tank contents to atmosphere will not exceed the i
c. 10CFR2O limits f or a release to an unrestricted area. (1.0) ~ i
'1 g
REFERENCE i i cREFERENCE PI Tech Specs, T.S. 3.9-9 OOOO59 GOO 4 ..(KA's) . 1
- h. (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) j
___-_j
i
. I i
Page 85 52 EMERGENCY AND ABNORMAL PLANT _ EVOLUTIONS 132h1 j i
)
)
ANSWER 5.04 (1.00) l
- ANSWER
- b. ...to provide a stable heat sink for decay heat removal. :(1.0) j REFERENCE
- REFERENCE EDP 1ES-0.34, Natural Circulation Cualdown with CRDM Fans, pg. 5. or 15.
ERG-HP Lackground Document for Natural Ci r'dul ati on . Cool down (ES-0.2), pg 28. ; OOOO15K307 ..(KA's) i l
)
ANSWER 5.05 (2.00)
)
- a. 3. ...ONE RHR pump must be operable and in operation, but may be taken out of service to f acilitate movement of f uel . (1.0) )
- b. 1. Suspend all operations in containment. (1.0)
REFERENCE PI Tech Spec's, pg. TS 3.8-1. 0000255003 ..(KA*s) ANSWER 5.06 (1.00)
- c. An overlap of one source range detector and one intermediate range detector will permit startup af the reactor. (1.0)
REFERENCE PI Tech Spec Table 3.5-2, pg i Of 2. System Description B9, Nuclear Instr umentation System, Figure B9-1 and pg. 38 of 39. OOOO32 GOO 3 ..(KA's) (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
,I Page 86. ]
E __gDESggBCy;8UD_8BNgRMAL__ PLANT _gVg(UTIONS 12231-i 1 ANSWER .5.07 '(2.00) OANSWER
- o. 4. Steam flow is equivalent ~ to rupture flow plus f eedwater flow. -(1.0) ~
- b. 1. a. size of tube failures- (1. 0) ' *
- b. . capacity of SI. system
- c. RCS cooldown rate REFERENCE eREFERENCE 1 ERG-HP, Background Document f or E-3, Steam Generator Tube Rupture, 'p. 4 1
thru 15. 000038A102 ..(KA's) ANSWER 5.08 (1.00) OANSWER
- a. 1, 5, 3, 6, 2, 4 (1.0)
REFERENCE GREFERENCE 1F-0, Critical Safety Function Status Trees OOOO74G012 ..(KA's) 1 ANSWER 5.09 (1.00) '
.l OANSWER j
- d. 1. S/G pressures - etable or decreasing L 2. RCS subcooling (based on core exit T/C's) - >~20 degrees F ;
- 3. RCS cold leg (WR) temperatures - at saturation for'S/G pressure .]
.I (1.0)
.i i
REFERENCE l GREFERENCE EDP 1ES-0.1, Reactor Trip Recovery, Attachment A : OOOO11A209 ..(KA's)~ l l (***** CATEGORY 5 CONTINUED ON NEXT'PAGE *****) .l
Page 87 5 EdgBygNCy_ANp_8pNgBd36_P69NT EVglyTIgNS 13_331 I j ANSWER 5.10 (2.00) cANSWER .f
- a. Site Area Emergency (0.5), Condi ti on #2, Loss of coolant accident with leak rate in excess of available pump capacity, OR on Condition
#19. (0.5) '
] 1
- b. Alert (0.5), Condition #13, Fuel damage accident with release of i radioactivity to' containment OR Condition 19, Conditions that warrant activation of TSC and nearsite EOF. (0.$)
REFERENCE CREFERENCE EPIP F3-2, pg. 11 of 61 and 49 or 61. OOOO11 GOO 2' OOOO36 GOO 2 ..(KA's) s ANSWER 5.11 (2.00) l
- o. 1. 1E-O OR Reactor Trip or Saf ety injection. (0.5)
- 2. F-0.1 OR Subcriticality CSF Status Tree (0.5)
- b. SI actuation will trip the MFW pumps. (1.0)
REFERENCE r Background Inf ormation f or FR-S.1, Response To Nuclear Power , I Gsneration/ATWS, pg. 1 of 4. OOOO29K312 ..(KA's) ANSWER 5.12 (1.50) CANSWER
- 1. 2 hour (0,5)
- 2. SGTR (0.5) >
- 3. 1.0 gpm (0.5)
REFERENCE oREFERENCE PI Tech Spec basis f or T.S. 3.1.D.1, pg. T.S. 3.1-12.
~
OOOO76 GOO 4 ..(KA's) (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
j 4 Page 88
,_7z__EMER@ENCY_6Np_ApNgRMAL PLANT EVOLUTIgLJE 122Z1- l
'1 f
ANSWER 5.13 .(2.00)
*ANSWEF: !
- a. NIS negative rote trip (1.0)
- b. 2.. a. Reduce reactor power to 80%. '
- b. Maintain delta I within. target band.
i (1.0) REFERENCE i OREFERENCE , cREFERENCE Oparating Procedure C5i Control Rod System, pg. 14 thru 15._ f OOOOO3 GOO 3 OOOOO3K103 ..(KA's) ) ANSWER 5.14 (1.50) oANSWER
- c. Power Range, Hi Flux Low Power Trip (0.5) ;
- b. Power Range hi Flux Trip (0.5)
- c. OT delta T Trip (0.5)
REFERENCE 3 I a.EFERENCE R Lesson Plan PB161L-OO4, pg. 3 of 14 thru 6 of 14. OOOOO1K103 ..(KA's) i i ANSWER 5.15 (1.00)
# ANSWER
.d. Containment Pressure r- 4 psig ,
RCS Subcooling based on T/C's = 30 degrees F .; AFW flow to intact S/G = 50 gpm l Narrow range level in intact S/G = 20% Pressurizer level = 10% - { RCS pressure is 0200 psig and stable ; (1.0) i l
-i j
(***** CATEGDRY 5 CONTINUED'ON NEXT'PAGE *****) l I
E __EdgBggtLCy_B!jD ABNORMAL _PL ANT _gVQLLITIQNS Page 89
.!?2Z1 REFERENCE CREFERENCE EDP 1E-1, Loss of Reactor or Secondary Coolant, pg. 7 of 15.
OOOO40A205 ..(KA's) ANSWER 5.16 (2.00) cANSWER (Procedure Number OR Procedure Title required for full credit) 1
- 1. 1E-0, Reactor Trip or Safety injectionh (0.5Q (O.
a.
- 2. 1ES-0.1, Reactor Trip Recovery. g{
- 3. 1ES-0.3A, Natural Circul ation Cooldown with CRDM Fans. (O.M) -
4 1EE-0.5 &, Naturol OrtutotiOO COodctWn u)WW cdoM 4&As. (o .ss)-
- b. Loss of charging (prior to loading onto the Diesel Generators) ( 1. 0 )
REFERENCE cREFERENCE EDP's Abnormal Procedure AB1, Lons of All Offsite Power, pg. 6 OOOO56A268 OOOOS6G012 ..(KA's) ANSWER 5.17 (1.00)
- ANSWER
- c. Lo-Lo Tavg and Safety injection Signal (1.0)
REFERENCE OREFERENCE System Description B-27, Main and Auxili ary Steam System, pg. 10 of 18 thru 11 of 18. 4 OOOO40K301 ..(KA*s) ANSWER 5.18 (1.00) j 1 cANSWER , prevent the pressurizer saf ety valves from lif ting f ollowing
- c. ...to a blackout. (1.0) ,
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
i
]
i
.Page 90 0._ EbsB9ENGX_8SD_8EU9BdBL_Eb6NI_EV969I1gNg l J
- ( 337. )
lJ REFERENCE !
,4 OREFERENCE System Description B-2EB, Auxilaary Feedwater System, pg. 4 of 21.
..(KA's) j OOOOO7K106 1
ANSWER 5.19 (2.00)
- ANSWER OANSWER l
- a. 1. Reactor Containment Vessel (0.5)
- 2. Shield Building (0.5) b.- 1.. a. Holds leakage f or decay.
- b. Dilutes leakage prior to rel ease.
(1.0) i REFERENCE OREFERENCE PI USAR, pg. 14.9-1. OOOO11A112 ..(KA's) I ANSWER 5.20 (1.50) 4 l j CANSWER cANSWER
- a. When the break flow is beyond the capacity of the charging pumps.
(0.50) (1.0) l
- b. 2. Low pressurizer pressure. .)
REFERENCE CREFERENCE q PI USAR, pg. 14.7-2 thru 14.7-4. OOOOO9A225 OOOOO9K304 ..(KA's) !
. t i
I (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
.- {
m
.* ,o Page 91 L.__EDEBEEd9.L699- ABU9Bd66_Ek6NI_EY969119BS 129Z4 1 l
l ANSWER 5.21 (3.00) GAMSWEh
- c. 2. If radiation level is 15 R/hr, Unit 2 shall be placed in cold' shutdown. (0.34)
- 4. Sample mixed bed demineralized for both influent and effluent. j (o.34)
- 6. Gurvey auxiliary building. (0.34)
- b. 1. Close containment purge valves. (0.33)
- c. 1. Close containment purge valves. (0.33)
- d. 3. Turn on all turbine b1dg roof ventilators. (0.33)
- 5. Check steam line radiation monitors. (O.33).
- a. 3. Turn on all turbine b1dg roof ventilators. (0.33)
- 5. Check steam 'line radiation' monitors. (0.337 REFERENCE Alcrm Res.ponste Index C47.47, Radiation Monitoring, Train A.
OOOO61 GOO 5 ..(KA's) i j i j i (***** END OF CATEGDRY 5 *****) 1
T I l Page 92 'I A __ELONI_SX9IEME_Is931_BND_ELBNI: WIDE _9ENEBIC BESE9BEID16111ES_lls31 1 ANSWER 6.01 '1.00)
)'
cANSWER Decrease in main feedwater temperature. (1.0) e. REFERENCE . CREFERENCE Westinghouse Rx Core Control For Large PWR, pg. 2-29 thru 2-3'O 035010K501 ..(KA's) I I s ANSWER 6.02 (1.00) 1 1 CANSWER ,
- a. ... constant (around 2000 gpm) AND this total flow is maintained by automatically controlling the RHRS heat exchanger flow bypass valve (CV-31237)." (1.0)
.1 REFERENCE OREFERENCE l
Oparating Procedure C-15, RHRS, pg. 2 System Description B-15, RHRS, pg. 13 of 28 i OOSOOOK402 ..(KA's) .i d ANSWER 6.03 (2.00) OANSWER
- c. 4. ...is unable to detect a stuck rod. (1.0)
- b. 3. . . .contains primary coils in series. (1.0) ,
REFERENCE CREFERENCE System Description B-6, pg. 4 of 14 and 5 of 14 014000K502 ..(KA's) (***** CATEGORY 6 CONTINUED DN NEXT PASE *****)
i l PLAST _gygIgtjg_j3Og1_6Sp_P($91-WJQg_ggNggig - Page 93 ! 6. j BESE99EID161IIES_ll331 i ANSWER 6.04 (2.00)
.I OANSWER aJk-Low level in the steam generator (1.0) .
- b. Place the pump control switch in " manual" (prior to draining the steam generator ) . (1.0)
REFERENCE ; OREFERENCE Northern States Power LER 87-006-01 061000K101 ..(KA's) ANSWER 6.05 (2.50) OANSWER GANSWER i
- o. 1) Steamline Isolation (0.75) l
- 2) Feedwater Isol ati on (0.75) 1
- b. 3. Rapid RCS cooldown OR Rapid RCS pressure increase while the !
RCS i s col d. (1.0) l J REFERENCE OREFERENCE 1E, Volume 2, F-0.4-Integrity, pg 1 of 3 System Description B-18C, Engineered Safeguards Features, pg. 6 of 36 ' 013OOOK416 ..(KA's) ANSWER 6.06 (1.00) OANSWER
- c. . . .the generator would produce voltage af ter starting locally if enough residual magnetism were present. (1.0) l REFERENCE
- REFERENCE System Description B-20.7, Emergency Diesel Generator, pg. 8 of 11.
System Description B-38A, Diesel Generators, pg. 35 or 37. 063OOOK301 ..(MA's)
.l
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)' ;
.]
i lj
.Page 94 6___ELONI_EXSIEUE_J2931_6BD_EL6BI: WIDE _GENEBIC
BEEE9BE1ElblI1ES_lls%1 ANSWER- 6.07 (1.00) CANSWER
- a. The condensete storage tank supply cross connect valves to the !
auxiliary f eedwater pumps must be tagged open. (1.0) REFERENCE i oREFERENCE l P1 Tech Spec 3.4-1 and Basis j 061000 GOO 6 ..(KA's)
-l ANSWER 6.08 (1.00)
OANSWER
- b. Lube oil pressure less than 6 psig on the #11 MFW pump will generate an automatic trip of the pump. (1.0) q REFERENCE !
- REFERENCE System Description B-28A, Condensate and Feedwater System, pg. 13 of 27 thru 14 of 27.
059000 GOO 7 ..(KA's) 1 J ANSWER 6.09 (1.00) ; cANSWER
- d. The containment spray "Stop-Normal-Start" control switch is in the 9
" NORMAL" posi ti on , and a "P" signal is received. (1.0) i REFERENCE ;
OREFERENCE i System Description B-1BD, Containment Spray System, pg. 7 of 16 thru B of 16. ! 022OOOA102 ..(KA's) , i i l. i (***** CATEGORY 6 CONTINUED DN NEXT PAGE *****)
.-----___--_---_-_--_a
Page 95 6z__EL8NI_gy@Iggg_Jgg31_6NQ_EL6NI;WlpE_ggyEBig EEEEQNSigibillgg_flg31 ANSWER 6.10 (1.00)
# ANSWER
- c. The SI test line isolation valves are closed to prevent e release of radioactive gases to the Auxiliary Building environment. (1.0)
REFERENCE CREFERENCE EDP Background Inf ormation f or 1ES-1.2, Transfer to Recirculation, pg. 2 of 6. 02602OK404 ..(KA's) ANSWER 6.11 (1.00) CANSWER
- o. The incore thermocouple are located over fuel assemblies in the core to measure f uel assembly outlet temperatures. (1.01 REFERENCE cREFERENCE System Description B-10, pg. 4 of 31.
017020K102 ..(KA's) ANSWER 6.12 (1.00) cANSWER i CANSWER
- c. Differential current (1.0) i REFERENCE !
GREFERENCE Leccon Plan B186L-OO4, pg. 30 l 064000K402
..(KA*s) l I
' i l l j i l i (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) j
1
,. 7 a
i; PLBSI_SYSIEUS_129Z-1_BSD_EbeUI: WIDE _@gNgBig Page'96 j 6. EEDE9BSIBILIIIEE_lisZ1 -1 j i i 1 1 l ANSWER 6.13 (2.00) ] l
- a. 2. S/G PORV's (0. 40) .
- 3. safety valves - (0. 40) ,
- b. 4. MSIV's (0.40) !
- 5. check valve immediately downstream of MSIV (0.40)
- c. 1, steam flow nozzle (0.40)
REFERENCE I OREFERENCE ' System Description B-27, Main and Auxiliary Steam System, pg. 3 of 18 through 9 of 18. 039000K406 ..(KA's) i
.i ANSWER 6.14 (1.00) l l
- a. ...enssre radiation at the surface remains belcw 50 mR/hr during fuel !
transfer. (1.0) REFERENCE , Lesson Plan PB187L-OO3,; Fuel Handling, pg. 11 of 84. 034000 GOO 6 ..(MA's) I 1 i ANSWER 6.15 (1.00) l
')
cANSWER
- o. A f uel supply of 70,000 gallons is available in the interconnected ;
storage tanks to supply one train of diesel generators for 14 days.- (1.0) REFERENCE 1 OREFERENCE , PI Tech Spec 3.7-3 and Basis 064000K608 ..(KA's) i i
-(*****. CATEGORY 6 CONTINUED ON NEXT PAGE.*****) l
- n .; .* ,e-Page 97 6- E6BBI_5XSIEDE_139%) AND_ELONI: BIDE _9ENEBIC BESE90EIE161IlES_Ilshi i
1 ANSWER 6.16 (1.00) CANSWER ) lc. . With the reactor half loaded with old fuel, containment integrity'(1.0) IS NOT required if RCS Temperature remains below 120 degrees F. REFERENCE l OREFERENCE PI Tech Spec 3.6-1 and Basi s 103OOOK303 ..(KA's) ,
'i ANSWER 6.17 (2.00) )
OANSWER
- a. Low (fire protection) header pressure (108 psig). (1.0) l
- b. Diesel Fire' Pump. (1.0) 1 REFERENCE cREFERENCE GREFERENCE
. System Description B-31A, Fire Protection System, pg. 18 of 54.
OB6000K402 ..(KA*s) ANSWER 6.18 (1.00) CANSWER
- d. When the pressurizer temperature reaches saturation, the temperature downstream of the PORV will. increase to the saturation temperature.
(1.0) REFERENCE OREFERENCE f Operating Procedure C1.2, Unit Startup Procedure, pg. 3 thru 8. 010000A104 ..(KA*s) (**a** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
f Page 98 dz__ PLANT SYSTEMS'(30%) AND PLANT-WIDE GENERig. BEEE9L4EIDILIIIES_ll231 l l l ANSWER 6.19 (1.00)
# ANSWER.
. . . maintain nearl y constant RCS water mass. (1.0) a.
REFERENCE . GREFERENCE Lesson Plan PB170L-OO6, Pressurizer Level Control System, pg. 28 of 43. OO2020K505 ..(KA's) ANSWER 6.20 (1.00) cANSWER
- b. An undervoltage condition of 75% rated voltage on Bus 11, with a 1/2 per bus logic, will trip the reactor in anticipation of a loss of coolant flow. (1.0).
REFERENCE cREFERENCE System Description B-8, Reactor Protection System, Table B-8-1. i 012OOOK402 ..(KA's) I 1 ANSWER 6.21 (1.00) '; cANSWER
- c. Reactor Vessel Dynamic Head indication is the actual differential. !
pressure across the reactor core and internals compared to the anticipated differenttel pressure f or normal, single-phase coolant flow. (1.0) REFERENCE GREFERENCE i System Description B-4B, RVLIS, pg. 3 of 11 thru 10 of 11. c i OO2OOOK107 ..(KA's) l 4 i l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
-j
r
]
Page 99 j 6 __ELONI_EYSIEDS_12931_8BD_EL6HI: WIDE SENEBig i BEEE9NelBILIIIES_ll231 ANSWER 6.22 (1.00) cANSWER-
- c. LOCA, Loss of secondary coolant, Steam generator t,ube rupture, and Rod ejection accident. (1.0)
)
. REFERENCE l rrREFERENCE l System Description B-1BB, ECCS, pg. 4,or 21 thru 5 of 21. OO6000 GOO 4 ..(KA's) ANSWER 6.23 (3.00) cANSWER s,
- o. 3. 10.0% (1.0) l
- b. 4. 15.0% (1.0) i
- c. 2. 7.5% (1.0)
REFERENCE l System Description B-27, Main and Auxiliary Steam-System, pg. 13 of 19. System Description B-7, Reactor Control System, pg.-15 of 54 thru 17 of 54. l 041000 GOO 7 ..(KA's) l I ANSWER 6.24 (1.00) i
- c. The Independent Verification can not be perf ormed by the first line i i
supervisor. (1.0) ! i REFERENCE Administrative Control Dir ective SACD 3.9, pg 3 of 16. l 194001K107 ..(KA's)
)
(*****' CATEGORY 6 CONTINUED ON NEXT.PAGE *****)
Page100 6z__ PLANT _gygTEMS_J29Z.l_AND_ PLAST-Wigg_gggggJg SEgEgyg1BILIT_I_Eg_J1331 ANSWER 6.25 (2.50) CANSWER GANSWER
- a. 1. Corrosion products (0.5)
- 2. Fission products '(0.5)
Perform an intentional crud burst each outage. (0.50)
- b. 1.
- 5. Minimi=e power' rate changes. (0.50) ~ '
- 6. Mini mize oxygen level s in ' the . RCS. . (0.50)
REFERENCE GREFERENCE Plant Safety, Section F2, Radiation Safety, pg. 3 thru 4. 194001K104 ..(KA's) ANSWER 6.26 (3.00)
- a. 2, 2,3,4 E Deduct 0.23 pts f or each missing .mswer
- b. 1, 3, 4 or each incorrect arswer given. If'no
- c. 1, 3, 4 currect answers given, -3.0 3
- d. 1, 3, 4 REFERENCE Plcnt Safety, Section F5, Fire Fighting, pg. 3 thru 5.
194001K116 ..(KA's) ANSWER 6.27 (2.00) .j q cANSWER
- o. 1. Area with a continuous dose rate of 3.5 mR/hr. (0.5% l (0.5) l
! b.- 3. Barricaded or roped off area.
- c. 5. Spent resin tank room. (0.5) ,
l (0.5) ? l d. 3. Barricaded or roped off area. l ; I i l i d 1 (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
.. s 1
- 6. PLANT _ SYSTEtiS_130%) AND PLANT-WIDE GENERIC Page101 RESPONSIBILITIES (13%1 REFERENCE 0 REFERENCE Plant Safety, Section F2, Radiation Safety, pg. 14 thru 16.
194001K103 i
..(KA's) 1 i'
ANSWER 6.28 (2.00) 0 ANSWER '
- 1. 'The tags are removed to allow testing of the equipment following maintenance. (1.0)
- 2. The tags are removed to allow other equipment activities to occur 1' (while the WR or precedure is active). (1.0)
REFERENCE CREFERENCE ~ Administrative Work Instruction, 5AWI 3.10.3, pg. 10 of 22 thru 19 or 22. 194001K102 ..(KA's) ANSWER 6.29 (2.00) CANSWER
- c. Control algae growth (in the pipes and components) (0.75)
- b. Inhalation of chlorine gas (0.75)
- c. An emergency shower or any other means of washing with water should be used (immediately). (0.5) '
REFERENCE i cREFERENCE D14, Chemical Handling, pg. 10 C30.1, Chlorine System, pg. 2 194001K111 ..(KA's) 1 (***** END OF CATEGORY 6 *****) (****u***** END OF EXAMINATION **********)
.* h
.a TEST CROSS REFERENCE Page 1 QUESILQM ___VALIJE BEFERENGE_
4.01 1.00 9000095 4.02 2.00 9000097 4.03 1.00 9000098 4.04 1.00 9000099 4.05 1.00 9000100 4.06 1.00 9000101 4.07. 1.00 9000102 4.08 1.00 9000103 4.09 1.00 9000104 4.10 1.00 9000105 4.11 1.00 9000106 4.12 2.00 9000107 4.13 1.00 9000090 4.14 1.00 9000091 4.15 1.00 9000092 4.16 1.00 9000093 4.17 1.00- 9000094 4.18 1.00 9000096 4.19 1.00 9000144 4.20 1.00 9000145 4.21 2.00 9000146 24.00 5.01 2.00 9000087 5.02 1.50 9000088 5.03 1.00 9000089 5.04 1.00 9000133 5.05 2. 0'O 9000134 , 5.06 1.00 9000135 I 5.07 2.00 9000136 9 5.08 1.00 9000137 l 5.09 1.00 9000138 i 5.10 2.00 9000139 5.11 2.00 9000140 ; 5.12 1.50 9000141 5.13 2.00 9000142 i 5.14 1.50 9000143 5.15 1.00 9000147 ; 5.16 2.00 9000148 l 5.17 1.00 9000149 ' , 5.18 1.00 9000150 ! 5.19 2.00 9000151 i 5.20 1.50 9000152 5.21 3.00 9000153 33.00 6.01- 1.00 9000083 6.02 1.00 9000084 . 6.03 2.00 9000085 l 6.04 2.00 9000086 G.05 2.50 9000108 ra rirA n ra Friferism i
.- )
,e & ;
d l 1 TEST CROSS REFERENCE Page 2
-I QUESILQH VALUE BEFERE_NCE.
6.08 1.00 9000111 , 6.09- 1.00 9000112 l' 6.10 1.00 9000113 6.11 1.00 9000114 6.12 1.00 9000115 6.13 2.00 9000116 l 6.14 1.00 9000117 . 6.15 1.00 9000118 1 6.16 1.00 9000119 I 6.17 2.00 9000120 .j 6.18 1.00 9000121 6.19 1.00 9000122 6.20 1.00 9000123 j 6.21 1.00 9000124 6.22 1.00 9000125 6.23 3.00 9000126 6.24 1.00 9000127 ' 6.25 2.50 9000128 6.26 3.00 9000129 6.27 2.00 9000130 'l' 6.28 2.00 9000131 6.29 2.00 9000132 43.00 ] 100.0
.l j
1 I
I - JASTER COPY U. S. NUCLEAR REGULATdRY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION REGION 3 FACILITY: Prairie Island 1 & 2 REACTOR TYPE: r'WR-WEC2 DATE ADMINISTERED: 89/06/I2 INSTRUCTIONS TO CANDIDATE: Use separate paper for the answers. Write answers on one i;ide only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the cuestion. The passing grade requires at least 70% in each category and a final grade of at l 1 east 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALU5 , CATEGORY 25.00 25.00 1. REACTOR PRINCIPLES (7%) ,
THERMODYNAMICS (7%) AND I COMPONENTS (11%) (FUNDAMENTALS ! EXAM) ) 27.00 ' 27.00 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS (27%) ) l 48.00 48.00 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC RESPONSIBILITIES (10%) 100.0 % TOTALS FINAL GRADE All work done on this examination is my own. I have neither given nor received aid. Candidate's Signature MASTER COP
i
. t.
- 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 2 i (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM) ]
]1 QUESTION 1.01 (1.00)
Moderator temperature coefficient becomes more negative from BOL to E0L. WHICH ONE (1) of the following statements describes the primary reason for this effect? (1.0) (a.) A decrease in the fuel to clad gap over core age which results in a decrease in the fuel temperature. j i (b.) A decrease in the boron concentration during core life j which results from fuel burnup and the production of tission ' fragment poisons. 4 (c.) Plutonium building over core age which results in more fissionable material being available to compete with boron atoms for neutrons. (d.) Plutonium building over core age causes hardening of the neutron flux which results in more fast neutrons available for fast fission and an increase in the fast fission factor. 1 I l 1.01 l ANSWER (1.00) 1 1 (b.) [+1.0) REFERENCE
- 1. Westinghouse, Reactor Core Control for Large Pressurized !
Water Reactors, Ch. 3, p. 3-28. { 192004K106 292004K110 292004K105 292004K101 ..(KA's) i (***** CATEGORY 1CONTINUEDONNEXTPAGE*****)
.. 1 I
'1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 3 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION 1.02 (1.00) WHICH ONE (1) of the following statements regarding fuel temperature I (doppler) coefficient (FTC) is correct? (1.0) i (a.) The magnitude of FTC will increase as reactor power increases.
~
(b.) FTC is always negative throughout e re life. (c.) FTC is a result of moderator temperature changes causing a , change in the resonance cross sections of.the fuel. (d.) FTC becomes more negative as fuel temperature increases. i ANSWER 1.02 (1.00) (b.) [+1.0] i REFERENCE
- 1. Westinghouse, Fundamentals of Nuclear Reactor Physics, Ch. 6, page 6-41 192004K102 ..(KA's) i i
(***** CATEGORY 1CONTINUEDONNEXTPAGE*****)
. o
- 1. REACTOR PRINCIPLES (7%'L THERMODYNAMICS Page 4 (7%)~AND COMPONENTS (13%) (FUNDAMENTALS EXAM) i l
QUESTION 1.03 (1.00) _ l During a reactor startup from an. initial Keff of 0.88, the first : reactivity addition caused count rate to increase from 10 counts
~l per second to 15 cow +s per second. The second reactivity addition ;
caused count rate to increase from 15 counts per second to 30 l
~
counts per second. WHICH ONE (1) of the following statements describes the relationship between the first and second reactivity additions? (1.0) i i (a.) The first reactivity addition was'the larger of the two. (b.) The second reactivity addition was the larger of the two. (c.) The first and second reactivity additions were approximately equal. (d.) There is not enough data to determine the relationship. ANSWER 1.03 (1.00) (a.) N [+1.0] REFERENCE
)
- 1. Westinghouse, Fundamentals of Nuclear Reactor Physics, Ch. 8.
i 192008K104 ..(KA's) (***** CATEGORY 1CONTINUEDONNEXTPAGE*****)
?, , b
- 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 5 ;
(7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM) QUESTION 1.04 (1.00) A reactor trip occurs from. full power equilibrium xenon conditions. Six (6) hourt later the reactor is brought critical at 10E-8 'Jys. : WHICH requirementsONE for (1)the of the nextfollowing(statements concerning two 2) hours is correct? (ASSUME rod motion ' reactor power is held constant at 10E-? amps) (1.0) (a.) Rods will have to be withdrawn since xenon will closely follow its normal production rate following a trip. l (b.) Rods will have to be inserted since xenon will closely , follow its normal decay rate following a trip. 1 (c.) Rods will have to be rapidly inserted since the critical i reactor will cause a high rate of xenon burnout. i (d.) Rods will have to be rapidly withdrawn since the critical reactor will cause a higher than normal production rate of I xenon. l ANSWER 1.04 (1.00) (a.) [+1.0] REFERENCE
- 1. Westinghouse, Reactor Core Control for Large Pressurized Water Reactors, Ch. 4, p. 4-11.
192006K107 ..(KA's) , (***** CATEGORY 1CONTINUEDONNEXTPAGE*****)
l n; i
' 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS- Page 6 .
(7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM) i c QUESTION- 1.05 (1.00)- The ratio of both Pu-239 and Pu-240 atoms.to U-235 atoms changes over core life. WHICH ONE-(1) of the following pairs of parameters is MOST affected by this change? (1.0) _, (a.) Moderator temperature coefficient and doppler coefficient. (b.) Doppler coefficient and beta. (c.) Beta and thermal neutron diffusion length. (d.) Themal neutron diffusion length and moderator temperature coefficient. ANSWER 1.05 (1.00) (b.) [+1.0] REFERENCE
- 1. Westinghouse, Reactor Core Control for Large Pressurized Water Reactors, Ch. 2, p. 2-13.
192003K107 192004K107 ..(KA's) QUESTION 1.06 (1.00) q WHICH ONE (1) of the following conditions results in.an INCREASE in' j available Shutdown Margin?. CONSIDER each case separately. -. ASSUME l the plant is operating at power with rods in AUTO co1 trol. (1.0) j (a.) RCS boron concentration decreases 50 ppe (b.) a single control rod mechanically binds at 200 steps 'l (c.) RCS Tavg decreases 5 degrees F j (d.) samarium concentration increases to its equilibrium value of 700 pcm (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)- i
P
, ' 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 7 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM) l ANSWER 1.06 (1.00)
(d.) [+1.0] i REFERENCE I
- 1. Westinghouse, Reactor Core Control for Large Pressurized Water Reactors, Ch. 7., p. 7-13.
i 192002K114 ..(KA's) QUESTION 1.07 (1.00) Which one of the following statements defines differential control rod worth? (1.0) (a.) Amount of reactivity change resulting from a unit change I in rod position. , I (b.) Total reactivity change resulting from a change in rod position over several units of rod travel. (c.) The difference in reactivity change resulting from movement of a center rod compared to an outer rod. (d.) The difference in a control rods worth at BOL and EOL. ANSWER 1.07 (1.00) (a.) [+1.0] REFERENCE
- 1. Westinghouse, Reactor Core Control for Large Pressurized Water Reactors, Ch. 6, p. 6-14.
192005K105 ..(KA's) (***** CATEGORY 1CONTINUEDONHEXTPAGE*****)
1 y - j l
- 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 8 ;
(7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM) j l l QUESTION 1.08 (2.00) i WHAT are FOUR (4) operating conditions that must be met to ensurt hot channel factor limits are maintained? (2.0) ] ANSWER 1.08 (2.00)
- 1. No misaligned control rods (+/- 12 steps) [+0.5]
- 2. Proper control rod sequencing [+0.25] and overlapping [+0.25]
- 3. Control rods maintained above to rod insertion limits [+0.5)
- 4. Delta-I (AFD) is maintained within the target band ' [+0.5]
d# REFERENCE
- 1. Prairie Island: Technical Specifications, p. 6 10-10.
1 193009K107 ..(KA's) I i 4
?
(***** CATEGORY 1 CONTINUE 0 ON NEXT FAGE *****)
.L
- 1. REACTOR PRINCIPLES-(7%) TilERM0 DYNAMICS Page 9 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION 1.09 (2.00) MATCH the valves listed below with the operational characteristic that best describes their use. (USE each operational characteristic only once) (2.0) VALVE TYPE OPERATIONAL CHARACTERISTICS
- a. Globe 1. Good flow control in.high flow, low pressure applications. Can cause water hammer.
- b. Gate 2. Used to prevent high system pressure.
Problems include leakage, incorrect lifting pressure, and failure to seat.
- c. Butterfly 3. Generally used in low flow,-high pressure applications wien precise flows are required.
- d. Check 4.- Used most often for throttling flow. ,
Capable of handling large pressure i differentials and can tightly shutoff ; flow.
- e. Relief 5. Generally used as on-off valves, small !
delta pressure when fully open, may ' need equalizing lines.
- f. Diaphragm 6. Generally used in low pressure and temperature applications, not normally i used for throttling.
- 7. Used to limit flow through a pipe, has ;
no moving parts. i j 8. Allows flow in only one direction, ! self opening and closing. ' l l l (***** CATEGORY 1CONTINUEDONNEXTPAGE*****)
i i
' 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 10 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM) ,
i ANSWER 1.09 (2.00)
- a. 4 '
- b. 5
- c. 1 )'
- d. 8
- e. 2 6
f.
)
[+0.33] each - I l REFERENCE
- 1. Westinghouse, Thermal Hydraulic Principles and Applications ;
to the Pressurized Water Reactor, Ch.10, p.10-62. j 193006K115 ..(KA's) j i QUESTION 1.10 (1.00) A system containing saturated vapor at 720 psig is leaking via a relief l valve. The pressure downstream of the relief valve is 15 psig. 1 ASSUMING ideal throttling characteristics, WHICH ONE (1) of the i following represents the condition of the fluid downstream of the relief valve? (1.0) . (a.) wet vapor /100% quality (b.) wet vapor /60% quality (c.) superheated vepor (d.) saturated liquid i ANSWER 1.10 (1.00) ; (c.) [+1.0] , 1 (***** CATEGORY 1CONTINUEDONNEXTPAGE*****)
' 1. REACTOR PRINCIPLES (7%) THERH0 DYNAMICS Page 11 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
REFERENCE
- 1. Westinghouse, Thermal-Hydraulic Principles and Appiteations to the Pressurized Water Reactor, Ch. 2, p. 2-48.
193004K115 ..(KA's) QUESTION 1.11 (1.00) WHICH ONE (1) of the following plant parameter changes v ali: vesult in operation closer to DNB7 (1.0) (a.) increased RCS Tavg (b.) increased RCS pressure (c.) increased RCS flow (d.) decreased reactor power ANSWER 1.11 (1.00) (a.) [+1.0] REFERENCE
- 1. Westinghouse, Thermal Hydraulics Principles and Applications to the Pressurized Water Reactor, Ch.13, pp. 23 and 24.
193008K105 ..(KA's) (***** CATEGORY 1CONTINUEDONNEXTPAGE*****)
l 4 6
- 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 12 l (7%) AND COMPONENTS (11%) (FUNDAMENTALS f.'XAM) l i
J QUESTION 1.12 (1.00) ASSUME a calorimetric power calculation is performed under each of the following conditions. WHICH ONE (1) of the following conditions would result in a CALCULATED power HIGHER than ACTUAL power? (Consider each separately) (1.0) (a.) Measured feedwater temperature is 10 degrees lower than actual feedwater temperature. 1 (b.) Measured steam generator pressure is 30 psig higher than actual steam generator pressure. l (c.) Measured feedwater flow is 1E5 lbm/hr lower than actual j feedwater flow. (d.) Measured feedwater header pressure is 30 psig higher than actual feedwater header pressure. 1 1.12 l ANSWER (1.00) (a.) [+1.0] REFERENCE ;
- 1. Prairie Island: Operating Procedure C41.4, pp. 7-12.1 ,
193007K106 ..(KA's) , I
'I l
l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
' 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 13 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION 1.13 (1.00) WHICH ONE (1) of the following conditions describes the indications that would be c5 served if a centrifugal pump were started and operated with its discharge valve shut? (ASSUME NO recirculation flow) (1.0) (a.) Higher starting current and lower running current. (b.) Lower starting current and lower running current. (c.) Higher starting current and higher running current. (d.) Lower starting current and higher running current. ANSWER 1.13 (1.00) (b.) [+1.0] REFERENCE
- 1. Westinghouse, Themal-Hydraulic Principles and Applications to the Pressurized Water Reactor, Ch.10, p.10-43.
191005K104 ..(KA's) QUESTION 1.14 (1.50)
- a. EXPLAIN the SINGLE basic principle of operation of stationary flow measuring devices. (0.5)
- b. STATE THREE (3) types of devices (no moving parts) that use this principle. (1.0) l
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
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- 1. REACTOR PRINCIPLES (7%i THERMODYNAMICS Page 14 (7%) AND COMPONENTS (1:L%)_(FUNDAMENTALS EXAM) f ANSWER 1.14 (1.50) ,
- a. Flow rate is proportional to the square root of the delta pressure (delta-P). [+0.5] ,
1
- b. 1. orifice
- 2. venturi
- 3. flow nozzle
- 4. delta-P across pipe bend ;g Any three (3) [+0.33] each I b hfactI bn(#
h tfwe SC l f'A t e # REFERENCE g g,;y
- 1. Westinghouse, Thermal Hydraulic Principles and Applications to Pressurized Water Reactors, Ch. 11., p. 11-17.
191002K105 ..(KA's) QUESTION 1.15 (1.50)
- a. DEFINE pump cavitation. (0.5)
- b. WHICH ONE (1) of the following sets of conditions are indications that a centrifugal pump is cavitating? (1.0)
(1.) fluctuating current, stable discharge pressure, excessive suction pressure. (2.) Fluctuating current, oscillating discharge pressure, stable suction pressure. (3.) Oscillating discharge flow,-control valve position, stable discharge pressure, 5 psig suction pressure. (4.) Excessive discharge pressure, relief valve open, normal suction pressure. (***** CATEGORY 1CONTINUEDONNEXTPAGE*****)
'1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 15 (7%) AND COMPONENTS (13%) (FUNDAMENTALS EXAM) . ANSWER 1.15 (1.50)
- a. Cavitation in a pump is the formation and subsequent collapse of vapor bubbles [+0.5].
- b. (2.) [+1.0)
REFERENCE
- 1. Westinghouse, Thermal Hydraulic Principles and Applications to the Pressurized Water Reactor, Ch. 10, p. 10-53.
191004K101 ..(KA's) QUESTION 1.16 (1.00) WHICH ONE (1) of the following MOST accurately describes the steam relief capacity of the Main Steam System? (1.0) (a.) The code safeties MUST be combined with the atmospheric dump vt41ves to have a capacity of 100%, with the relief settings from 1005 psig to 1020 psig. (b.) The code safeties MUST be combined with the atmospheric dump valves to have a capacity of 100%, with the relief settings from 1075 psig to 1129 psig. (c.) The code safeties DO NOT have to be combined with the atmospheric dump valves to have a capacity of 100%, with relief settings from 1005 psig to 1020 psig. (d.) The code safeties DO NOT have to be combined with the atmospheric dump valves to have a capacity of 1005, with relief settings from 1075 psig to 1129 psig. ANSWER 1.16 (1.00) (d.) [+1.0] (***** CATEGORY 1CONTINUEDONNEXTPAGE*****)
' 1. -REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 16 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
REFERENCE
- 1. Prairie Island: System Description B-27, Main and' Auxiliary System, p. 7.
191001K101 ..(KA's) QUESTION 1.17 (1.00) WHICH ONE (1) of the_following phrases correctly completes the sentence: "An undercompensated ion chamber compensates out ...."? (1.0) (a.) more neutrons which gives a lower signal than anticipated. (b.) less neutrons which gives a higher signal than anticipated.
-(c.) less gama radiation which gives a higher signal than anticipated.
(d.) more g6mma radiation which gives a lower signal than anticipated. ANSWER 1.17 (1.00) (c.) [+1.0] REFERENCE
- 1. Prairie Island: Lesson Plan P8184L-002, Obj. 8.
191002K118 ..(KA's) l QUESTION 1.18 (1.00) Concerning CVCS DEMINERALIZERS: -
- a. WHY is' letdown flow limited to 90 gpm? (0.5)-
- b. WHY is letdown temperature maintained below 140 deg F7 (0.5)
(***** CATEGORY 1CONTINUEDONNEXTPAGE*****)
'1. REACTOR PRINCIPLES (7%? THERMODYNAMICS Page 17' (7%) AND COMPONENTS (1?,%) (FUNDAMENTALS EXAM)
ANSWER 1.18 (1.00)-
- a. The cation demineralized has a rated flow of 90 gpm (re enerative heat exchan er flow capacity)(prevent -
resnchanneling). [+0.5
- b. to protect ion exchanger resins [+0.5]
REFERENCE
- 1. Prairie Island: Lessc.) Plan, P8172L-001A, p. 14.
- 2. Prairie Island: System Description B12A, p. 16.
- 3. Prairie Island: Operating Procedure C12, p. 6.
191007K104 ..(KA's) QUESTION 1.19 (1.00) WHICH ONE (1) of the following effects will occur as a result of increased fouling of the mein condenser tube bundles while at 100% power?' (ASSUME circulating water flow and steam flow rates . remain constant) .(1.0) (a.) Condenser heat rejection will increase (b.) Circulating water outlet temperature will decrease (c.) Condensate depression will increase (d.) Condenser hotwell temperature will decrease l ANSWER 1.19 (1.00) (b.) [+1.0] (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
' . 1REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 18 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
REFERENCE
- 1. Westinghouse Themal-Hydraulic Principles and Applications to i the PWR, Ch. 9, p. 9-19.
191006K112 ..(KA's) ] i QUESTION 1.20 (1.00) WHICH ONE (1) of the following statements describes how the steam generator pressure indication will compare to actual steam generator pressure if containment pressure is increased to an abnormally high pressure? (1.0) (a.) Indicated steam generator pressure will be HIGHER than actual steam generator pressure. (b.) Indicated steam generator pressure will be LOWER than actual steam generator pressure. j i (c.) Indicated steam generator pressure will be EQUAL to actual steam generator pressure. (d.) Indicated steam generator pressure will fail as is. ANSWER 1.20 (1.00) 1 (c.)MrQ. [+1.0) l REFERENCE I
- 1. Westinghouse Thermal Hydraulic Principles and Applications to PWR, Ch. 11, p. 11-11.
191002K111 ..(KA's) Pro 39ztr l 1 (***** CATEGORY 1CONTINUEDONNEXTPAGE*****) i _ .:_---_-_-_D
'1 REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 19 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION 1.21 (1.00) WHICH ONE (1) of the following would result in the temperature detector output signal failing HIGH7 (1.0) (a.) An open circuit for an RTD. (b.) A short circuit for an RTD. (c.) An open circuit for a thennoccuple. (d.) A short circuit for a thermocouple. ANSWER 1.21 (1.00) (a.) [+1.0] REFERENCE l
- 1. Westinghouse, Thermal-Hydraulic Principles and Applications to l the Pressurized Water Reactor, Ch.11, p.11-4. '
191002K114 ..(KA's) ! (***** CATEGORY 1CONTINUEDONNEXTPAGE*****) - _ _ - _ _ - _ _ _ _ _ _ _ . - - .~
' 1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 20 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION 1.22 (1.00) WHICH ONE (1) of the following statements correctly describes why density compensation of the main steam line flow measurement is necessary? (1.0) (a.) Differential pressure across the flow venturi in the steam line is proportional to the volumetric flow rate. Volumetric flow rate is compensated with the fluid density to provide mass flow rate. (b.) Differential pressure across the flow venturi in the steam line is inversely proportional to the volumetric flow rate. Volumetric flow race is compensated with the fluid temperature to provide mass flow rate. (c.) The temperature of the steam lines is proportional to the volumetric flow rate. Volumetric flow rate is compensated with the fluid temperature to provide mass flow rate. (d.) The temperature of the steam lines is inversely proportional to the volumetric flow rate. Volumetric flow rate is compensated with the fluid density to provide mass flow rate. ANSWER 1.22 (1.00) (a.) [+1.0] REFERENCE
- 1. Prairie Island: Lesson Plan P8174L-006, p. 20.
191002K102 ..(KA's) l (***** END OF CATEGORY 1*****)
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- 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 21 (27%) j 1
1 QUESTION 2.01 (1.00) Pressurizer pressure transmitter (1P-429) fails high. STATETWO(2) ' control system responses that would identify the failed channel as the contrciling channel. (1.0) ANSWER 2.01 (1.00)
- 1. Both pressurizer spray valves open. [+0.5]
- 2. Pressurizer heaters turn off. [+0.5] I J
REFERENCE
- 1. Prairie Island: Instrument Failure Guide C-51, IC51.1, Section IP-429, p. 3.
000027A215 ..(KA's) l QUESTION 2.02 (1.00) l WHICH ONE (1) of the following is the setpoint for the low condenser vacuum turbine trip? (1.0) ! (a.) 20" Hg (b.) 22" Hg (c.) 23" Hg (d.) 24.5" Hg ANSWER 2.02 (1.00) (a.) [+1.0] l l l (***** CATEGORY 2CONTINUEDONNEXTPAGE*****). l
- 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 22 (27%)
l REFERENCE
- 1. Prairie Island: System Description B-22A, p. 32. l 000051A202 ..(KA's) 1 QUESTION 2.03 (1.50)
With Unit I at 100% power Train A DC is lost. WHAT are THREE (3) automatic actions (other than annunciators) that l occur? (1.5)- ANSWER 2.03 (1.50) -) l l 1. reactor trip
- 2. turbine trip
- 3. transfer of DC power for buses 11(21) and 12(22) to the opposite Unit's battery train l
- 4. transfer of computer inverters to alternate source ,
Any three (3) [+0.5] each; max. [+1.5] REFERENCE
- 1. Prairie Island: Opes t.ng procedure C20.9, pp. 16 and 17.
000058G010 ..(KA's) QUESTION 2.04 (1.50) Seal injection has been lost to an RCP and attempts to restore it have failed. WHATTHREE(3)immediateoperatoractionsare required prior to the lower bearing water temperature reaching 200 deg F7 (ASSUMEreactorpowerat30%) (1.5) l 1 1 (***** CATEGORY 2CONTINUEDONNEXTPAGE*****)
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'2. ' EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 23 ,
(27%)- l I l ANSWER 2.04 (1.50) : 1.- Trip reactor [+0.5] i
- 2. Stop affected RCP [+0.5)
- 3. Close number 1 seal leakoff isolation valve,for affected RCP (CV-31426 or CV-31427) [+0.5) i REFERENCE
- 1. Prairie Island: Operating Procedure 2C3.1, p. 7. 1 1
000015A208 000015G010 ..(KA's) i I QUESTION 2.05 (1.00) I WHICH ONE (1) of the following situations would require a controlled plant shutdown? (ASSUME 100% reactor power.) Consider each case separately. (l'.0) j (a.) " Operational Basis Earthquake" annunciator is received and l confirmed by the graphic recorder. ' 1 (b.) Rod position indicator channel out of service. (c.) The Daily Leak Rate Surveillance has just been completed . and indicates an increase in Reactor Coolant leakage to ! 7.0 gpm identified leakage. i (d.) One containment spray pump out of service for one (1) hour. ANSWER 2.05 (1.00) l (a.) [+1.0] (***** CATEGORY 2CONTINUEDONNEXTPAGE*****) i
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- 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Paga'24 1 I
(27%)- REFERENCE
- 1. Prairie Island: AB-3, p. 3. . l
- 2. Prairie Island: Technical Specification 3.1.9, 3.3.3, and 3.10-6. i 000003G011 000007G008 000007G011' ..(KA's) 1 QUESTION 2.05 (1.50) l AccordingtoEmergencyOperatingProcedureE-0,"Rea:torTripor Safety Injection, WHAT are THREE (3) means of identifying a ruptured 3' steam generator? (1.5)
ANSWER 2.06 (1.50)
- 1. High SGB radiation. [+0.5]
- 2. High condenser air ejector radiation. [+0.5] ]
- 3. High radiation for SG steamline monitor. [+0.5]
l REFERENCE i
- 1. Prairie Island: ORP 1E-3, " Steam Generator Tube Rupture," 1 P. 3.
000037A206 ..(KA's) 4 QUESTION 2.07 (1.50) a.. With Unit 1 at 100% power, an uncontrolled insertion of a single control rod occurs. STATE TWO (2) operator actions that can be performed to stop the uncontrolled insertion.' (1.0) i
- b. WHAT action is required if TWO control rods are involved? (0.5) i
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
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~ 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 25 (27%)
l ANSWER 2.07 (1.50) l
- a. 1. Place rod control system in manual. [+0.5]
- 2. Open rods lift coil disconnect switch. (manual '
reactor trip) [+0.5]
- b. Manually trip reactor. [+0.5]
REFERENCE
- 1. Prairie Island: C-5, 8,4, Uncontrolled Insertion of an RCCA,
- p. 20. ,
j 000005G010 000005A203 ..(KA's) 1 QUESTION 2.08 (1.50) I Due to a fire in the control room it becomes necessary to evacuate. STATE THREE (3) actions to be performed prior to leaving the control 6.b room. (>4 ) 1 I J ANSWER 2.08 (1.50)
- 1. trip reactor ,
- 2. close MSIVs
- 3. close pressurizer PORV block valves
- 4. verify turbine trip Any three (3) [+0.5] each; max. [+1.5]
REFERENCE l
- 1. Prairie Island: Plant Safety Procedure F-5, Appendix B,
- p. 4. -
000068K312 ..(KA's) (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
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~ 2. -EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 26 (27%)
1 QUESTION 2.09 (1.00) j The following indications are recognized in the control room, in the followin reactorat 100%)g sequence. (ASSUME no operator action and-
- 1. charging flow decreases I
- 2. pressurizer level decreases 1
- 3. letdown isolation and pressurizer heaters off
- 4. pressurizer level increases toward high level reactor trip point WHICH ONE (1) of the following failures has occurred? (1.0)
(a.) Controlling pressurizer level channel failed low. (b.) Auctioneered Tavg failed high. (c.) Controlling pressurizer level channel failed high. (d.) Reference pressurizer level failed to no-load value. i l
. ANSWER 2.09 (1.00)
(c.) [+1.0] REFERENCE
- 1. Prairie Island: Lesson Plan P8170L-006, Terminal Objective. '
000028A212 ..(KA's) (***** CATEGORY 2CONTINUEDONNEXTPAGE*****)
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~ 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 27- ,
I (27%) d i QUESTION 2.10 (2.00) I
- a. STATE FOUR (4) indications of a steamline break inside containment that could ALSO be symptoms of a LOCA. (Prior to a reactor trip) (1.0)
- b. STATE THREE (3) indications that would only be symptoms !
of a steam break, and that could be used to differentiate - between the two events. (1.0) l 2.10 I ANSWER (2.00) ,.
- a. 1. RCS pressure decrease 7- bd Pd [;F8 m IU^ .f i
- 2. pressurizer level decrease '
- 3. VCT level decrease I
- 4. containment pressure .
- 5. containment temperature increase
- 6. containment humidity increase Any four (4) [+0.25] each; max. [+1.0] , ,
- b. 1. high steam flow with low steam pressure [+0.33] 4 [# ' "# *) '
- 2. Iow steam pressure [+0.33]
- 3. containment airborne radiation [+0.33] $ $G fnrpn knu.'c '
lac of C. hhl dm [/ *
- i nur aje/ ded O~
REFERENCE 7.
- 1. Westinghouse, Thermal-Hydraulic Principles and Applications to Pressurized Water Reactors, Ch.14, p. 32. .
000011A213 000040A203 ..(KA's) (***** CATEGORY 2CONTINUEDONNEXTPAGE*****)
- 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 28 (27%)
QUESTION 2.11 (3.00) ANSWER the following questions concerning IFR-S.1, " Response To Nuclear Power Generation /ATWS."
- a. HOW is a reactor trip verified? (STATE FOUR (4) acdons andexpectedresponses.) (2.0)
- b. WHAT FOUR (4) contingency actions are contained in the response not obtained column (RNO) for verifying a turbine trip? (1.0)
ANSWER 2.11 (3.00)
- a. 1. rod bottom lights [+0.25), lit [+0.25]
- 2. 0.25] and bypass breakers [+0.25] open
- 3. reactor trip [ in rod position +dicators [+0.25), zero [+0.25]
- 4. neutron flux [+0.25], decreasing [+0.25]
- b. 1. manuallytripturbine[+0.25] '
- 2. verify all turbine control valves closed [+0.25]
3.
- 4. manually close MSIVs close and control valves[+0.25 bypass valves [+0.25))
REFERENCE a
- 1. Prairie Island: FRP 1FR-S.1, p. 3.
000029G010 ..(KA's) QUESTION 2.12 (1.00) Per E-0, " Reactor Trip or Safety Injection," WHAT conditions require tripping the RCPs given the following parameters. ,
- 1. Containment pressure is 4.0 psig
- 2. Containment radiation is 1000 mRen/hr. (1.0)
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
- 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 29 (27%)
ANSWER 2.12 (1.00)-
- 1. at least one SI pump running (SI flow). [+0.5]
- 2. RCS pressure less than 1250 psig [+0.5]
3 An (2c5 e..\Lankas nd bu^ In%dI L, % ersh REFERENCE
- 1. Prairie Island: E0P 1E-1, " Loss of Reactor or Secondary Coolant," p. 3.
000011A103 ..(KA's) QUESTION 2.13 (3.00)
- a. WHAT THREE (3) contingency actions are required in FRS.1,,
"ATWS" to establish emergency boration if nonnal boration of the RCS can NOT be established? (1.5)
- b. WHAT is the basis for emergency w ration flow being '
limited to 12 gpm? (1.0)
- c. HOW is emergency boration flow Iimited to 12 gpm? (0.5) i ANSWER 2.13 (3.00)
- a. 1. Shift the running boric acid transfer pump [+0.25] to fast [+0.25] ,
- 2. OpenBASTrecircvalve(CV-31195to50%[+0.5]
- 3. Openemergencyborationvalve(MV-32086) [+0.5]
- b. Potential for loss of seal irjection flow (due to acid solidification in seal injection throttle valves). [+1.0]
- c. Throttle the emergency boration flowmeter outlet valve (VC-11-58) [+0.5
(***** CATEGORY 2CONTINUEDONNEXTPAGE*****)
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' 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 30 (27%)
I REFERENCE
- 1. Prairie Island: FRS.1, "ATWS," p. 3. ,
000024K302 ..(KA's) QUESTION 2.14 (1.00) WHICH ONE (1) of the following Emergency 0)erating Procedures are j you required to COMPLETE even if a red pati exists in the Critical :) Safety Function Status Trees? (1.0) ! (a.) LOSS OF ALL AC POWER RECOVERY WITH il REQUIRED (ECA 0.2). (b.) LOSSOFREACTORORSECONDARYC00LANT(E-1). (c.) LOSS OF ALL AC POWER (ECA 0.0). (d.) STEAM GENERATOR TUBE RUPTURE (E-3). ANSWER 2.14 (1.00) (c.) [+1.0] REFERENCE
- 1. Prairie Island: Emergency Operating Procedures.
000055G012 000011G012 000038G012 000054G012 ..(KA's) , (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
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- 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 31 i (27%)
i QUESTION 2.15 (1.00) WHICH ONE (1) of the following is NOT'an innediate operator action I for 1ECA-0 0, " Loss of All AC Power?" (1.0) l 1 (a.) Verify AFW flow to all intact SGs. (b.) Verify reactor trip. !
)
(c.) Check.if RCS is isolated. l (d.) Attempt to restore power to any safeguard bus. l l ANSWER 2.15 (1.00) (d.) [+1.0] l REFERENCE
- 1. Prairie Island: 1ECA-0.0, " Loss of All AC Power," pp. 3 and 4.
000055G011 ..(KA's) l QUESTION 2.16 (2.50) Step 5 of the IlEEDIATE actions in E-0, " Reactor Trip or Safety Injection," has you verify Safeguards Component alignment. STATE FIVE (5) indications./ components that you have to verify to complete this step. (INCLUDE in you answer the re condition / position of each component / indication.) quired (2.5) 1 i (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
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- 2. EMER'GENCY AND ABNORMAL PLANT EVOLUTIONS Page 32 l
(27%) ANSWER: 2.16 (2.50)-
- 1. Category I doors [+0.25] - closed [+0.25]
2. Ched closedOps log][+0.25] [+0.25 for any) (withinsix(6 ventilation openings that must be minutes)
- 3. SI not ready lights [+0.25] - OFF (not-lit) [+0.25]
(withexceptions)
- 4. SI active lights [+0.25] - ON (lit) -[+0.25] (with exceptions) ;
q
- 5. Containment isolation lights [+0.25] - ON (lit) [+0.25] q (withexceptions) h REFERENCE
- 1. Prairie Island: E-0, " Reactor Trip or Safety Injection,
- p. 5.
000007K301 ..(KA's) QUESTION 2.17 (2.00)
~With Unit 1 at 100% power, a loss of offsite power occurs.
STATE FOUR (4) conditions by which you can verify that natural circulation exists. INCLUDE parameters AND expected indication. (2.0) s i 4
)
l -i j l
-l
(***** I CATEGORY 2CONTINUEDONNEXTPAGE*****)'
' 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 33 (27%)
ANSWER 2.17 (2.00)
- 1. Steam pressure stable or decreasing.
- 2. RCS hot leg temperature stable or decreasing.
- 3. Core exit thermocouple stable (..- decreasi,g. ,
- 4. RCS cold leg temperature at' saturation temperature for steam pressure.
- 5. RCS subcooling based on core exit thermocouple greater than 20 deg F.
Any four (4) [+0.25] each parameter, [+0.25] each indication; max. [+2.0] REFERENCE
- 1. Prairie island: 1ES-0.1, " Reactor Trip. Recovery," Attachment A.
- 2. Prairie Island: Lesson Plan P8161L-009, Obj.'1.
000056K101 ..(KA's) l (*****ENDOFCATEGORY 2*****)
" 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERI_C Page 34 RESPONSIBILITIES C0%)
QUESTION 3.01 (2.00) i Procedure FR-Z.1, " Response to High Containment Pressure," requires that hydrogen concentration be monitored.. ;
- a. WHAT is the primary source of hydrogen in containmen't during a major loss of coolant accident? (0.5)
- b. High containment hydrogen concentration increases the potential for an explosion and resultant pressure spike.
l WHY is this explosion a concern? (0.5)
- c. WHAT component is used to reduce hydrogen concentration while you are in this procedure? (0.5) l l
- d. IDENTIFY TWO (2) means for monitoring hydrogen concentration i in containment. (0.5) 'j i
l 1 ANSWER 3.01 (2.00) ;
- a. Zirconium-water reaction. [+0.5]
- b. Potential challenge to containment integrity. [+0.5] l
- c. Hydrogen recombiners. [+L ]
- d. 1. hydrogen recorder [+0.25]
- 2. gas analyzer [+0.25] (grab sample)
REFERENCE
- 1. Prairie Island: 1FR-Z.1, " Response to High Containment Pressure."
- 2. Prairie Island: Operating Procedure C-39.4, p. 2.
l l 3. Prairie Island: B-19, TC-87-001, Section 3.118. 028000K503 ..(KA's) i (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
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' 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC' Page 35 RESPONSIBILITIES D0%)
i QUESTION 3.02 (2.50). The SI test line to RWST isolation valves are interlocked with ! the containment sump "B" isolation valves.
- a. WHAT interlock must be satisfied to open the containment ' ,
sump "B" isolation valves? -(1.0) l
- b. STATE the THREE (3) bases for this interlock. (1.5)
I ANSWER 3.02 (2.50)
- a. One of the SI test lines to RWST isolation valves (MV-32202(8826A)) or MV-32202(88268)) must be shut.- [+1.0]
- b. 1. Loss of coolant back to the RWST. [+0.5]
- 2. Airborne radioactive release via the RWST. [+0.5]
- 3. Contamination of the RWST by water from sump "B". !
j [+0.5] REFERENCE
- 1. Prairie Island: System Description B-18A, p. 19. l 006000K409 ..(KA's)
QUESTION 3.03 (2.00) STATE FOUR (4) in)uts into RVLIS that provide for density i compensation of t1e indicated vessel level readings. (2.0). ! ) (***** CATEGORY 3CONTINUEDONNEXTPAGE*****)
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- 3. PLANT SYSTEMS (?B%D AND PLANT-WIDE GENERIC Page 36 RESPONSIBILITIES 00%)
ANSWER 3.03 (2.0@ h
- 1. RCS wide rangeGT temperature. [+0.5]
- 2. Wide' range RCS pressure. [+0.5]
- 3. Leveltransmitter(delta-Pcell). [+0.5]
- 4. Reference leg temperature (capillary tube temperature).
[+0.5] REFERENCE
- 1. Prairie Island: System Description B-48, pp. 7 and 8.
002000K503 ..(KA's) QUESTION 3.04 (1.00) STATE FOUR (4) ECCS components supplied by the component cooling water system. (USE redundant components only once) (1.0) ANSWER 3.04 (1.00)
- 1. RHR pumps
- 2. RHR heat exchangers
- 3. containment spray pumps
- 4. safety injection pumps
[+0.25] each REFERENCE
- 1. Prairie Island: Lesson Plan P872-002, p. 5 008000K102 ..(KA's)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
'3 PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 37 RESPONSIBILITIES (10%)
QUESTION 3.05 (2.00) ASSUME Unit 1 is at 100% power. Power range nuclear instrumentation channel N-41 fails high.
- a. WHAT FOUR (4) procection/ control circuits are affected by this channel failure? (1.0)
- b. Control rods step in as a result of this failure. WHY does this occur? (1.0)
ANSWER 3.05 (2.00) e reIc.ds*[h**."'E"g1l
- 1. a. OT delta-T r e:25, circs f-0"e al b. OP delta-T [M # ,-
oN c. power range high f' lux trip [+fhf5] low h"A P"P 2f ca'g d. power range high flux rod stop [4.25] 9 P't b r a 4t. f r <,p k- tv$s4G* rs k +iop 2. Mismatch rods to stepbetween turbine in to reduce the power and'[+0.5 mismatch reactor
. p]ower [+0.5] causing REFERENCE
- 1. Prairie Island: IC51.1, p. 33.
- 2. Prairie Island: Technical Specification 2.3-1.
- 3. Prairie Island: Lesson Plan P81841-002, Obj. 5.
015000K301 015000K405 ..(KA's) QUESTION 3.06 (2.00)
- a. WHAT TWO (2) signals will actuate Containment Spray System?
(INCLUDEcoincidenceandsetpoint). (1.0)
- b. WHAT TWO (2) benefits result from the addition of sodium hydroxide (NaOH) to the Containment Spray System? (1.0)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
- 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 38 RESPONSIBILITIES (10%)
ANSWER 3.06 (2.00)
- a. 1. Containment pressure signal at 23 psig [+0.25] (Hi-Hi pressure) 3 sets 1 of 2 coincidence [+0.25]
- 2. Manual [+0.25] 2 of 2 coincidence [+0.25]'
- b. 1. To control. containment sump pH [+0.5]
- 2. Enhance the removal of iodine [+0.5]
REFERENCE
- 1. Prairie Island: System Description B-18D, pp. 5 and 7.
026000K402 ...(KA's) QUESTION 3.07 (1.00) ; With UNITthe will cause 1 main at 100% power, feedwater WHICHvalve regulating ONE'(1) of the following(conditions to OPEN7 ASSUME initial valve direction) (1.0)
]
I (a.) An inadvertent SI actuation. (b.) Controlling steam generator level transmitter fails low. (c.) Controlling first stage pressure transmitter fails low. (d.) Controlling steam generator pressure transmitter fails low. l ANSWER 3.07 (1.00) (b.) [+1.0] l (***** CATEGORY- 3 CONTINUED ON NEXT PAGE *****) ;
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' 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 39 RESPONSIBILITIES (10%) ;
i REFERENCE
- 1. Prairie Island: Lesson Plan P8174L-006, p. 1. Obj. 6.
- 2. Prairie Island: System Description B-28a, p. 16.
059000K104 ..(KA's) QUESTION 3.08 (1.00) WHICH ONE (1) of the following statements is CORRECT concerning the intermediate range compensated ionization chambers? (1.0) (a.) The boron lined chamber is sensitive to neutron and gama I radiation, while the unlined chamber is sensitive only to gama rays. ' (b.) The compensated ion chamber is designed to remove the gama signal only at high reactor power levelt. (c.) Pulse height discrimination is used to compensate for ionization caused by alpha and gama. (d.) Gamas which penetrate the detector's lead shielding will cause ionization which results in a voltage pulse. ANS.TR 3.08 (1.00) (u.) [+1.0] REFERENCE
- 1. Prairie Island: Lesson Plan P8184L-002, pp. 20 and 21. -
015000K501 ..(KA's) (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
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.. j .'3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 40 RESPONSIBILITIES (10%)
J QUESTION 3.09 (1.00) l WHICH ONE (1) of the following is a Technical Specification limiting -) condition for operation (LCO)? (1.0) I (a.) Unidentified leakage .5 gpm (b.) Containment pressure 1.75 psig. (c.) Boric Acid Storage Tank 2100 gallons. (d.) RWST Tank 150,000 gallons. I 1 1 ANSWER 3.09 (1.00) (d.) [+1.0] , REFERENCE
- 1. Prairie Island: Technical Specifications pp. 3.1-9, 3.6- ,
3, 3.2-1, and 3.3-1. 002000G011 006000G011 ..(KA's) l QUESTION 3.10 (2.00) STATE THREE (3) automatic signals that will initiate a Safety Injection signal. (INCLUDEsetpoints) (2.0) ANSWER 3.10 (2.00)
- 1. Pressurizer pressure [+0.33] < 1815 psig [+0.33]
- 2. Containment pressure [+0.33] 2 4 psig [+0.33] :
- 3. Steam line pressure [+0.3.i] < 500 psig [+0.33]
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
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RESPONSIBILITIES D0%) { REFERENCE ;
- 1. Prairie Island: System Description, Section B-18A, p. 10.
013000K101 ..(KA's) .) i QUESTION 3.11 (2.00)
- a. WHAT is the purpose of the Number 1 seal bypass valve on the Reactor Coolant Pump (RCP)? (1.0)
- b. WHAT are the minimum and maximum pressure for opening the' -
Number 1 seal bypass valve? (1.0) ANSWER 3.11. (2.00) ]
- a. Provides additional flow to radial bearing for cooling. [+1.0]
- b. minimum pressure - 100 psig maximum pressure - 1000 psig [+[0.5))
+0.5 REFERENCE
- 1. Prairie Island: System Description, S-3.
- 2. Prairie Island: Operating Procedure C-3.
003000G007 003000G010 003000A110 ..(KA's) 4 (***** CATEGORY 3CONTINUEDONNEXTPAGE.*****)
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.'3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 42 l RESPONSIBILITIES (10%) I i
QUESTION 3.12 (3.00) ANSWER the following questions concerning the RHR System.
- a. WHAT TWO (2) interlocks are associated with the RCS to RHR pump suction isolation valves, MV-32164(32230) and MV-32165(32231)?
(INCLUDE setpoints) (control switches in AUTOMATIC) (1.0) a I
- b. During shutdown cooling operation, TWO (2) of the RCS to RHR i' umpsuctionisolationvalves(MV-32164(32230)andMV-32165 p(33231)) are open with breaker power removed. WHY is breaker power removed? (1.0)
- c. WHAT TWO (2) interlocks are associated with the RHR heat exchanger to SI pump suction valves (MV-32206 and MV-32207)?
(INCLUDEsetpoints) (1.0)- ARSWER 3.12 (3.00)
- a. 1. RHR hot leg isolation valves (MV-32164 and MV-32165) cannot be opened [+0.25] unless RCS pressure < 425 psi [+0.25].
- 2. RHR hot leg isolation valves (MV-32164 and MV-32165) close automatically [+0.25] if RCS pressure increases -
to 600 psig [+0.25]. .
- b. Prevent single instrument failure [+0.5 from causing loss of suction to RHR(s) (loss of shutdown)][+0.5].
- c. 1. The valves (MV-32206 and MV-32207) cannot be opened [+0.1]
unless the associated RHR pump disch* 4 rge pressure [+0.2] is < 210 psig [+0.2]. 1 1 2. l The unlessvalves (MV-32206 the associated and MV-32207) SI pump cannot suction isolation be [+0.2 valve opened [+0.1] ] l from the RWST is shut [+0.2]. l l i (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
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- 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 43 RESPONSIBILITIES (10%)-
REFERENCE
- 1. Prairie Island: System Description B-15, pp. 15, 16 and 17.
005000K407 ..(KA's) QUESTION 3.13 (2.00)
- a. STATE THREE (3) signals that would cause the motor-driven l auxiliary feedwater pumps to start automatically. (Control switch in automatic) (1.5)
- b. WHAT additional signal would start the turbine-driven auxiliary feedwater pump. (Controlswitchinautomatic) (0.5)
ANSWER 3.13 (2.00)
- a. 1. safety injection signal [+0.5] 4
- 2. low-low level in either steam generator (13%) [+0.5]
- 3. loss of both main feedwater pumps [+0.5]
- b. tlndervoltage condition on both bus 11 and bus 12. [+0.5]
REFERENCE
- 1. Prairie Island: System Description B-288, pp. 7 and 11.
061000K402 ..(KA's) i l l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
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' PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 44 RESPONSIBILITIES (10%)
QUESTION 3.14 (2.00) J MATCH the following permissives in COLUMN A with the respective ) function in COLUMN B. (ONE (1) answer only) (2.0) COLUMN A COLUMN B
- a. P-2 1. Allows blocking the source range Hi flux trip.
- b. P-4 2. Automatically unblocks low power trips.
- c. P-6 3. Reactor trip breakers open. ,
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- d. P-7 4. > 10% impulse. pressure which feeds P-7.
- c. P-8 5. Allows blocking of pressurizer high pressure trip.
- f. P-9 6. Automatically de-energizes SR high volts.
- g. P-10 7. Enables auto rod withdrawal.
- h. P-13 (unit 2 8. Automatically unblocks the turbine only) trip / reactor trip signal.
- 9. Automatically unblocks power range Hi flux trip.
- 10. Unblocks the single loop loss of flow trip.
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ANSWER 3.14 (2.00)
- a. 7.
- b. 3.
- c. 1.
- d. 2.
- e. 10. .
- f. 8.
- g. 6.
- h. 4. j
[+0.25] each
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REFERENCE
- 1. Prairie Island:
012000K610 ..(KA's)
)
QUESTION 3.15 (1.00) i WHICH ONE (1) of the following emergency diesel protective i trip inputs is active during a safety injection? (1.0). l 1 (a.) crank case high pressure .I (b.) reverse current (c.) differential current (d.) low lube oil pressure l ANSWER 3.15 (1.00) (c.) [+1.0] J (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
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REFERENCE ;
- 1. Prairie Island: Lesson Plan P8186L-004, p. 30.
064000K402 ..(KA's) l QUESTION 3.16 (2.00)
- a. WHY does re-aligning of a control bank A rod cause a rod control system urgent failure alarm? (1.0)
- b. WHY isn't an urgent failure alarm received when re-aligning a control bank B rod? (1.0)
I ANSWER 3.16 (2.00)
- a. Rod control senses an error because there is demand for both groups of rods in Bank A to move [+0.5], but all lift coils are' disconnected for one of the groups [+0.5]
- b. Bank B only contains one group of rods and one of them still has the lift could connected, so no alarm. [+1.0]
REFERENCE
- 1. Prairie Island: P8184L-005, Terminal Objective, .C5.
001010K605 ..(KA's) QUESTION 3.17 (1.00) The spent fuel storage racks are designed to prevent the possibility of inadvertent criticality. WHAT TWO (2) design features provide this reactivity control? (1.0) i (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
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" 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 47 RESPONSIBILITIES (10%)
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i 4 ANSWER 3.17 (1.00)
- 1. Physical separation (by the 9.5 inch centers). [+0.5]
- 2. Each storage tube has a layer of neutron absorbing material.
[+0.5] REFERENCE
- 1. Prairie Island: System Description B-17, p. 30.
033000K405 034000G004 ..(KA's) QUESTION 3.18 (2.00) A malianction occurs to the Charging Pump Controller in AUTO, and you are required to place that controller in MANUAL.
- a. DESCRIBE the steps to place controller in manual. (1.5)
- b. HOW is the pump speed adjusted while in manual? (0.5)
I 1 ANSWER 3.18 (2.00) j 1
- a. Transfer from AUTO to MANUAL:
- 1. Move transfer switch to AUTO-BAL [+0.5]
- 2. Adjust manual control knob for zero deviation [+0.5] !
- 3. Move transfer switch to MANUAL [+0.5]
- b. Pump speed is adjusted with the manual control knob. [+0.5] ,
REFERENCE e
- 1. Prairie Island: lesson Plan P8170-006, p. 22.
011000A404 ..(KA's) (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
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l QUESTION 3.19 (1.00) Unit ) has just tripped and the steam dump 'Is not functioning. l
- a. WHAT is the preferred means of automatically controlling primary system temperature in this condition? (0.5)
- b. At WHAT level will pressurizer level be maintained WITHOUT operator action? (0.5)
ANSWER 3.19 (1.00) '! I
- a. steam generator PORVs [+0.5] .
1
- b. 25.85%(+/-2%) [+0.5]
REFERENCE ;
- 1. Prairie Island: lesson Plan P8170 LOO 6.
039000A204 039000K402 011000A104 ..(KA's) QUESTION 3.20 (2.00) Safety Injection has been inadvertently actuated.
- a. STATE THREE (3) conditions and/or actions required to reset the SI signal. (1.5)
- b. After resetting a SI signal, WHAT action must be performed to allow automatic actuation of SI? (0.5) l
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' 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 49 RESPONSIBILITIES (10%)
1 ANSWER 3.20 (2.00) l l
- a. 1. 90 see time delay relay timed out [+0.5]
l 2. P-4 signal present (reactor trip breakers open) '[+0.5] )
- 3. BothSIreset(A:Btrains)pushbuttonsdepressed [+0.5] I l
l b. reactor trip breakers must be closed (P-4 signal cleared) [+0.5] REFERENCE
- 1. Prairie Island: Lesson Plan P8180L-006, pp. 40 and 6 .
013000K401 ..(KA's) ; QUESTION 3.21 (1.00) j l
- a. WHAT component in the CVCS system is used to reduce high ;
lithium concentration? (0.5) J
- b. WHAT component in the CVC5 system is used for RCS oxygen ,
control? (0.5) ANSWER 3.21 (1.00)
- a. cation demineralizers [+0.5]
- b. volume control tank [+0.5]
REFERENCE
- 1. System Description, B12A, pp. 12-14. ,
004000K504 004010K512 004000G007 ..(KA's) (***** CATEGORY 3 CONTINUE 0 ON NEXT PAGE *****) ;
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l QUESTION 3.22 (2.00) MATCH the following RCS Control System and Reactor Protection System inputs to the correct setpiont. (ONE (1) answer each) (2.0) RCS Control System and Reactor Protection System Setpoint
- a. backup heaters energize 1. 1800 psig
- b. SI block permissive 2. 1815 psig
- c. PORVs open 3. 1900 psig
- d. spray valves full cpen 4. 2000 psig
- a. hi pressure reactor trip 5. 2200 psig
- f. low pressure reactor trip 6. 2210 psig
- g. low pressure SI trip 7. 2310 psig i h. safety valves lift 8. 2335 psig l
- 9. 2375 psig
- 10. 2385 psig
- 11. 2485 psig
- 12. 2500 psig i
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'l ANSWER 3.22 (2.00) ;
- a. 6. :
- b. 4. 1
- c. 8. !
- d. 7. I
- e. 10. -
- f. 3. l
- g. 2.
- h. 11. I
[+0.25] each REFERENCE
- 1. Lesson Plan P8170L-005, Obj. 2.
010000A302 ..(KA's) i QUESTION 3.23 (1.50) ; According to SWI-0-2, " Shift Turnover," STATE $IX (6) information items that must be passed on to the incoming operator. (1.5) j ANSWER 3.23 (1.50)
- 1. Current Plant status
- 2. System status
- 3. Surveillance procedures outstanding
- 4. Maintenance in progress or planned
- 5. Plant operation evolutions in progress
- 6. Plant operation evolutions planned
- 7. Technical Specification related items
- 8. New operational procedures
- 9. New administrative procedures .
- 10. New administrative instructions. i Any six (6) [+0.25] each; max. [+1.5]
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REFERENCE
- 1. Prairie Island: SWI-0-2, pp. 8 and 9.
194001A105 ..(KA's) QUESTION 3.24 (1.00) If you are in a 100 mrad / hour gama field (QF-1) for 45 minutes, you would receive WHICH ONE (1) of the following doses? (1.0). (a.). 45 mrem (b.) 75 mrem (c.) 450 mrem (d.) 750 mrem ANSWER 3.24 (1.00) (b.) [+1.0] REFERENCE
- 1. 10CFR20.
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~~~WEpdNsisiLYffE$"(? 3M)AND PLANT-WIDE GEriERI
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QUESTION 3.25 (1.50) ENTER the 10CFR20 exposure limits (Rem) for a radiation worker in the blank spaces provided. (ASSUME no exposure extensions have been issued) QUARTERLY LIMIT
- a. whole body (0.5)
- b. skin (0.5)
- c. extremities (0.5)
ANSWER 3.25 (1.50)
- a. 1.25 rem .
- b. 7.5 rem
- c. 18.75 rem
[+0.5] each PEFERENCE
- 1. Prairie Island: Plant Safety Manual F2, p. 18.
i 194001K103 ..(KA's) l QUESTION 3.26 (1.00) . According to Prairie Island's Access Control Procedure, STATE l THREE (3) required actions you must perform before entering a Controlled Area. (1.0) . l
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- 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 54 I RESPONSIBILITIES (10%) ,
1 ANJWER 3.26 ()' l
- 1. read 'the. RWP ( A).33]
- 2. record the RWP [+0.165] number on your access control card [+0.165]
- 3. record yout dosimeter reading [+0.33]
REFERENCE l l
- 1. Prairie Island: Plant Safety Manual F2, pp. 22 and 23. l 194001K10S ..(KA's) 00ESTION 3.27 (1.00)
WHAT are the TWO (2) methods available that will determine if a diesel generator fuel oil day tank level column isolation valve requires independent verification following repositioning? (1.0) .
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ANSWER 3.27 (1.00) 1. Display)of the IVL on the Plant Information Computer (Sperry [+0.5] ;
- 2. Hard copy of the IVL in the control room. [+0.5]
adbd'At[ $1 fo8 00 ^ 3 Ske (th ew f (;Cgg h 9 f da bec k l g,f,f-bbM REFERENCE , 1. Prairie Island: Administrate ve Work Instruct;on SAWI 3.10.2, Independent Verification Of Components List, p. 9. 194001K101 ..(KA's) l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)-
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- 3. ' PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 55 j RESPONSIBILITIES (10%) q i
QUESTION .3.28 (1.50) I In accordance with F2, " Radiation Safety," - WHAT are THREE (3) precautions that MUST be taken wher:ever entries are made into an oxygen deficient atmosphere (< 16% by volume)? (1.5) ANSWER 3.28 (1.50)
- 1. Standby person available (to assist in an emergency).
- 2. self-contained breathing apparatus (SCBA) (Scott air pack)-
- 3. Communications between respirator wearer and standby person.
- 4. Person (working in the area) should be equipped with a safety ')
hamess and safety lines (or equivalent provisions). Any three [+0.5] each; max. [+1.5] REFERENCE .
- 1. Prairie Island: F2, " Radiation Safety," p. 26.
194001K113 ..(KA's) QUESTION 3.29 (1.50) .; STATEtheFOUR(4)levelsofemergencyclassifications, l according to F3-2, Emer 1 RANK in order of INCREAgency Plan Implementing' Procedure. ING severity. (1.5) {
- i. s )
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- 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 56 ,
RESPONSIBILITIES (10%)
. ANSWER 3.29 (1.50)
- 1. Unusual Event
- 2. Alert
- 3. Site Area Emergency (Site Emergency)
- 4. General Emergency
[+0.25] each; [+0.5] for correct order REFERENCE
- 1. Prairie Island: Emergency plan Implementing Procedure, F3-2, pp. 1 and 2.
194001A116 ..(KA's) QUESTION 3.30 (1.50) STATE the TWO (2) conditions that allow the temporary removal of HOLD and SECURE cards and temporary restoration of equipment. (1.5) ' ANSWER 3.30 (1.50)
- 1. Tags removed to allow testing of equipment (M .75]
(followingmaintenance)
- 2. Tagsremovedtoallowotherequipmentactivitiestooccur[+0.75] l (whiletheWRprocedureisactive) 6 l
REFERENCE
- 1. Prairie Island: AWI 3.10.3, pp. 18 and 19. j 194001K102 ..(KA's)
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+ # TEST CROSS REFERENCE Page 1 QUESTION ,VALUE REFERENCE 1.01 1.00 9000319 1.02 1.00 9000320 1.03 1.00 9000321 1.04 1.00 9000322 1.05 1.00 9000323 1.06 1.00 9000324.
1.07 1.00 9000337 1.08 2.00 9000325 1.09 2.00 9000326 1.10 1.00 9000338 * ' 1.11 1.00 9000339-1.12 1.00 9000340 1.13 1.00 9000327 1.14 1.50 9000328 1.15 1.50 9000329 1.16 1.00 9000330 1.17 1.00 9000331 1.18 1.00 9000332 1.19 1.00 9000333 1.20 1.00 9000334 1.21 1.00 9000335 1.22 1.00 9000336 25.00 2.01 1.00 9000341 2.02 1.00 9000342 2.03 1.50 9000343 2.04 1.50 9000344 2.05 1.00 9000345 1 2.06 1.50 9000346 2.07 1.50 9000347 2.08 1.50 9000348 2.09 1.00 9000349 2.10 2.00 9000350 2.11 3.00 9000351 2.12 1.00 9000352 2.13 ?.00 9000353 2.14 1.00 9000354 2.15 1.00 9000355 2.16 2.50 9000356 2.17 2.00 9000357 27.00 3.01 2.00 9000365 3.02 2.50 9000366 3.03 2.00 9000367 3.04 1.00 9000368 3.05 2.00 9000369 3.06 2.00 9000370 3.07 1.00 9000371 3.08 1.00 9000372 3.09 1.00 9000373 3.10 2.00 9000374
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9 TEST CROSS REFERENCE Page 2 ) l 00ESTION VALUE REFERENCE 3.11 2.00 9000375 ) 3.12 3.00 9000376 1 l 3.13 2.00 9000377 3.14 2.00 9000378 1 i 3.15- 1.00. 9000379 3.16 2.00 9000380 l 3.17 1.00. 9000381 .j 3.18 2.00 9000382 . y 3.19 1.00 9000383 ; 4 3.20 2.00 9000384 I 3.21 1.00 9000386 3.22 2.00 9000387 3.23 1.50 9000358 . l 3.24 1.00 9000359 l l 3.25 1.50 9000360 l 3.26 1.00 9000361 3.27 1.00 9000362 3.28 1.50 9000363 i 3.29 1.50 9000364 3.30 1.50 9000385
...... q 48.00
#m9999 100.0 !
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