ML20236B562
| ML20236B562 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 01/17/1989 |
| From: | Moorman J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20236B561 | List: |
| References | |
| 50-348-OL-88-01, 50-348-OL-88-1, NUDOCS 8903210162 | |
| Download: ML20236B562 (343) | |
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ENCLOSURE 1 EXAMINATION REPORT :50-348/0L-88-01 i
Facility Licensee':
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Alabama Power Company
'600 North 18th Street
' Birmingham, AL 35291-0400 Facility Name: J. M. Farley Nuclear Plant Facility Docket No.- 50-348 and 50-364 .
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- Written examinations.and operating tests were administered at the J. M. Farley Nuclear Plant'near Ashford, Alabama.
Chief Examiner: //./V2syv+. ( /- /4 89 Jame H. Moorman', III Date Signed Approved by: [ / /7![)
John F. Munro', Chief /Dat6 Signed i Operator Licensing Section 1 !
Summary:
Examinations on Novmeber 14-18, 1988.
Written examinations and 0perating tests were administered to seven candidates; six of.whom passed.
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Based on the results described above, three of three R0's passed and three of four SR0's passed.
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REPORT DETAILS
- 1. Facility Employees Contacted:
- R.' Hill, Assistant General Manager-0perations
- W.-Lee, Plant Instructor-Nuclear R. Lulling, Plant Instructor-Nuclear C. McLean, Plant Insrtuctor-Nuclear
- J. 0sterholtz, Manager-0perations ,
- L. Stinson, Assistant General Manager-Plant Support l
- B. Vanderbye, Plant Instructor-Nuclear i
- R. Wiggins, Supervisor-0perations Training
- L. Williams, Training Manager .
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- Attended Exit Meeting j
- 2. Exeminers:
- J. Moorman, III, Region II M. Morgan, Region II T. Guilfoil, Sonalysts, Inc. j F. Victor, Sonalysts, Inc. i
- Chief Exariner ;
- 3. Examination Review Meeting At the conclusion of the written examination, the examiners provided I R. Wiggins, with a copy of the written examination and answer key for review. The NRC Resolutions to facility comments are listed below,
- a. R0 Exam i (1) Question 1.02b l NRC Resolution: Ccmment accepted. This part of the question i
, has been deleted and the point value of the question lowered to 1.5 points.
(2) Question 1.08 NRC Resolution: Comment accepted. The answer will be changed as recommended by the facility. 3 (3) Question 1.11 i
NRC Resolution: Comment not accepted. The assumption that delta T remains constant in this situation is contrary to information provided on page 27 of FNP lesson plan OPS-40301C l (0PS-52101C), Steam Generator. Since other references were not provided by the facility to support their contention, no change i to the answer key is warranted. !
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f g '(4) Question 1.13 j .J NRC Resolution: Comment not accepted. The System curves will (
be graded according'to the answer key since-the knowledge tested
'is: an application of the facility material and not simple!
! memorization of~ facts. ; .[ ,
(5) Question 1.14 <
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' NRC Resolution: Comment accepted. The answer key will be ,
changed as recommended by the facility. '-
y 4 (6)' Question 12.03 ] j NRC Resolution: , Comment accepted. . .'The answer key ' will be 9, =
- changed as recommended'by the facility.
'(7)'Ouestion2.15 {
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'NRC Resolution: Comment accepted. The answer key has ben j, changed as ' recommended . by the f acility. Part d. of the .
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answer has been deleted. and the question value - has been -t changed to 1.50 points.
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, .(8) l Question 3.01 / ;
NRC Resolution: Comment accepted. 'The answer key Mill be changed to delete ' ' Low Main Steam Line . Pressure' from the b.,
part of the answer. The point value of the question,has bet:n lowered to 2.00' points.
(9) Question 3.05 NRC Resolution: Comment not accepted. The reduced or low. . ,
current aspect of the current hold placed on the moveable and stationary grippers is indeed part of the automatic actions'per the facility's own annunciator response procedure. No change to the answer key.
1 (10) Question 3.10 1 NRC Resolution: Comment accepted. .The taswer key will be 'i changed as recommended by the facility. ,
t (11) Question 3.12 l NRC Resolution: Comment acknowledged. The answer key'and -
the facility recommended answer are equivalent. The facility (
recommended answer will be added to the answer key for j clarification.
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f (12) Question 4.05 i
-NRC Res'olution: Comment acknowledged. The appropriate valve iy numbers will be added to-the answer key and accepted for full credii;.
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'(13) Question 4.09/7.14-4 '
Comment accepted. The question will be deleted y '[ NRC Resolution:
from trne exam.
(14) Question 4.10 I NRC Resolution: Comment accepted. The answer key will be l changed as recommended by the facility.
I (15)' Question 4.11 7 NRC Resolution: Comment ' not accepted. The question deals
, ,, with ' recognition' of the final ,STP schedule and not the process of scheduling, which is an SR0 function.
1 (16) Question 4.13 I NRC Resolution: Comment-accepted. The' typographical error in ..
t the answer key will be corrected as recommended by the facility. !
(17) Question 4.15
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- . ' NRC Resolution: Comment acknowledged. Reasonable answers 1
" I will be acceptable for full credit.
- b. SRO Exam o (1) Question 5.03
.NRC Resolution: Comment not . accepted. The question stated "Given the following absolute values of reactivity change over the 3 days:"... Rods = []6938 pcm and "[ SELECT THE CORRECT VALUE]". l Since the question stated a value of reactivity for rods, it was unnecessary to "use a ballpark value of 1000 pcm for the most reactive rod". Additionally, if that assumption was made, the candidate would most likely be unable to select the correct 6 answer as directed in the question. At that point, the candidate could have asked for clarification from the proctor. No change to the answer key. .
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4 (2) Question 5.07 NRC Resolution: Comment not accepted. The question did not state that the rod withdrawal occurred at 2386' cps and thus i a potentially critical status. The question indicated that the rod withdrawal occurred during the approach to criticality.
It should also be noted that no candidate indicated that "no correct answer exists". No change to the answer required.
(3) Question 5.20 NRC Resolution: Coment not accepted. Evidence was not provided by the facility to prove that 100% mixing occurs in the core during this transient. No change to the answer key.
(4) Question 6.10 NRC Resolution: Comment accepted. As a result of inaccurate reference material provided by the facility, parts D. and F.
will be deleted form the exam. Point values for the remainder of the answers will be increased to .45 points. However, it should also be noted that the question does not require the memorization of penetration sizes. It requires that an operator have an understanding of the relative size and location for RCS loop penetrations.
(5) Question 6.11 NRC Resolution: Comment not accepted. It has not been substantiated that this controller condition / indication is one which an operator need not understand. The question stated that the master pressure-controller fails at 43%. The facility has not justified the plausibility of their assumption. No change to the answer key.
(6) Question 7.10 NRC Resolution: Comment partially accepted. The facility indicates that the upender being moved to the spent fuel pool and the gate valve being shut are valid assumptions. These assumptions are not inherent to the question. General Emergency will remain the accepted answer to Part A. However, Alert will be accepted for full credit if these assumptions are clearly stated. The answer to Part B will not be changed since conditions provided in the question correlate directly to conditions specified in EIP-18, paragraph 3.2.3.
(7) Question 7.11 NRC Resolution: Comment accepted. " Manual 1/2" was deleted from the answer key and remaining part B required responses l were increased to 0.1 5 each.
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(8) Question;7.12 '
NRC Resolution: Comment accepted. Answer B.1. A will be 1 deleted from the answer key and the point value of .the question I will be lowered to 2.5 points.
(9) Question 8.03 NRC Resolution: Coment not accepted. The information provided to support the comment addresses the operation of. manual valves only and not MOV's. No change to the answer key. .1 (10) Question 8.04 NRC Resolution: Comment accepted. The answer key will be changed'as recommended by the facility.
(11) Question 8.14 NRC f<esolution: Comment partially accepted. The Part B of the question question asks how long operation may continue before 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of penalty deviation accumulates, not how long operation may continue. In accordance with TS 4.2.1.2, one hour is the correct answer. The answer key will be amended to also accept the facility's recommendation for Part 0 as an alternative answer.
(12) Question 8.16 NRC Resolution: Comment accepted. The raswer key was changed to accept the answer recommended by the facility for full credit.
- c. Further review of the answer key has resulted in the following changes:
(13) Question 7.09 Part B of the answer has been changed to delete 'then proceed to step 5' to be more reflective of the answer solicited by the question. The point values of the other parts were increased accordingly.
(14) Question 8.10 The answer has been expanded to include 'Whenever recorder charts are replaced' to cover all operational situations, j (15) Question 8.15 Point value distributions have been changed to be more consistent with the other allowable response to the question.
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- 4. Exit Megting I At the conclusion of the site visit the examiners met with representatives of the plant. staff to discuss the results of the examination. q Training material initially provided to the NRC for preparation of examina-tions did not contain a substantial amount of the information required by 10CFR55.41 and- 10CFR55.43 to be covered' in examinations. The licensee )
was cautioned that future inadequate submittals could result in ~ delay-or !
postponement of examinations as stated in the- examination notification letter. j One procedural weakness was noted as a result of candidate responses to i question 7.02 concerning turbine operation. Graphs in FNP-50P-1-28.1- 1 provided for determining time ~ limitations for turbine load changes and acceleration rate can be easily misinterpreted. This lead two.of the senior operator candidates to calculate non-conservative times for turbine operation.
There were no generic weal.nesses noted.during the oral examination.
The cooperation given to the examiners and the effort to ensure an atmosphere in the control room conducive to oral examinations was also noted and appreciated.
The licensee did not identify as proprietary any of the material provided to or reviewed by the examiners.
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l' U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: FARLEY REACTOR TYPE: PWR-WEC3 DATE ADMINISTERED: 88/11/14 EXAMINER: REGION II CANDIDATE: /H45 E/?.
INSTRUCTIONS TO CANDIDATE: 1 Uca separate paper for the answers. Write answers on one side only.
Stcple question sheet on top of the answer sheets, Points for each qucation are indicated in parentheses after the question. The passing grcde requires at least 70% in each category and a final grade of at locst'S0%. Examination papers will be picked up six (6) hours after the examination starts.
% OF
-CATEGORY % OF CANDIDATE'S CATEGORY -i VALUE TOTAL BCORE VALUE CATEBORY 'l 30.00 25.00 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 30.00 25.00 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 30.00 25.00 3. INSTRUMENTS AND CONTROLS 00 25.00 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 120.00 % Totals Final Grade 4 All work done on this examination in my own. I have neither given nor received aid.
Candidate's Signature e __
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules applys l
- 1. Cheating on the examination means an automatic denial of your application cnd could result in more severe penalties. ]
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil only to facilitate legible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the examination. ,
l S. Fill in the date on the cover sheet of the examination (if necessary). 1
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of each cection of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer sheet.
- 9. Number each answer as to category and number, for example, 1.4, 6.3.
- 10. Skip at least three lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTIDN AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not c1 car as to intent, ask questions of the examiner only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed. l 1
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- 10. When you complete your examination, you shall:
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- o. Assemble your examination as follows:
(1) Exam questions on top.
(2) Exam aids . figures, tables, etc.
(3) Answer pages including figures which are part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of the paper that yeau did not use for answering the questions. r
- d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked. 3 t
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EQUATION SHEET s f a na , v = s/t
, ' u = as s = v,e + kac 2 Cycle afficiency = E** , ""I E = sc -
a = (vg - v,)/t l KZ = mv 2 At A
y g,v g, ,c ,
kg A . 4, PE = agh w = 6/t A " la 2/t ig = 0.693/tg ,
W
- v4P '
AZ t (eff) = (t,')(ts) -
711Am
- i (Cg + tb)
Q=$ CAT I . g .-IX e P ,
o Q = UAAT I. , *VX
'Pvr = Wg 5 I= I,to -X/ M P=P 105URit). TVt = 1.3/u !
P=P o et /T HVI. e 0;693/u SUR = 26.06/T ~
T = 1.44 DT SCR = S/(1 - K,gg)
/1-- * ,c)
SUR = 26 g CR g = S/(1 > K,gg,)
T = '(t*/o ) + {(gi ',)/x,g,p] 1(1 ~ Kaff)1 CR2 II ~ Eeff)2 7 = g*/ (, . p M = 1/(1 - K,gg) = CR t/CRO I" ~ 8)! eff 8
M = (1 . E,gg)q/(1 . g.gg) 8 " IIsff"I)I aff * #eff/Eeff a=Q.
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p= ( t*/TK,gg -) + [I/(1 + 1,ggt )] 1* = 1 x 10 seconds P = I$v/(3 x 1010) 1 0.1 sec ds ~l I = Na -
Idgg=1d22 WATER PARAMETERS Id =Idl g
1 gal. = 8.345 lba 2 R/hr = (0.5 CE)/d (=eters) 1 gal. = 3.78 liters R/hr = 6 CE/d (feet) -
1 ft = 7.48 gal. MISCELI.ANEOUS CONVERSIONS .
3 Density = 62.4 lbm/f t 1 Curia = 3.7 x 10 dps O Density = 1 gm/cm 1 kg = 2.21 1ha Heat of var ori:stioni = 970 Etu/lbm I hp = 2.54 x 10 BTU 3
/hr Hest of fusien = 144 Btu /lbre 1 N = 3.41 x 106Btu /hr '
l' A t m = 14. 7 P s i = 2 9.'9 in . 1 3 1 Btu = 778 f t-lbf
*- *** '- 2 i f t .' o = 0.'4 3 35 'Ib f /in g . inch = 2.54 cm T = 9/5 C + 32
'C = 3/9 ('r . 32) l___________
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- 1. PRINCIPLES OF NUCt_ EAR POWER PLANT OPERATION, PAGE 2 IfigftMODYNAMICS. HEAT TRANSFER AND_ELUID FLOW GUESTION 1.01 (1.50)
TRUE or FALSE a). During reactor startup, "true criticality" has occurred when the-fission chain reaction is stable, T-AVE is constant.and a positive SUR is being maintained with minimal rod motion.
Hb ) Once a positive startup rate is ar. helved, the reactor is then made slightly subcritical (by rod insertion) to ensure supercritical core conditions will never be met.
.c) Theoretically, if the neutron source could be removed with Keff exac;1y equal to ONE (1) - and Intermediate Range indication at 10E(-)8 amps - the neutron level would drop to a lower value and then remain constant.
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'i. ' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 3 i IBERfiO]WWtICS. HEAI_TRANSEER...ANIL FLUID FLOW L.
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-QUESTION i.02 +2,40) -
- CHOOSE / CIRCLE THE APPROPRIATE WORD WHICH COMPLETES THE FOLLOWING SENTENCES
I Givcn caximum heat generation rate Peaking Factor = ------------------~~--------
average heat generation rate L
c) The (AXIAL / RADIAL / LOCAL) peaking factor would be the one which accounts for the variance.in power up and l down the vertical axis of the core.
l Jyr- The (OPERATING / DESIGN / ACCIDENT) peaking factor would be the one which exists.in a " worst case" condition which i might be expected at any time in plant life.
I l c) The (AXIAL / RADIAL / LOCAL) peaking factor would account ]
for variances in power produced at various distances I out from the vertical axis of the core. l l
d) The (OPERATING / DESIGN / LOCAL)-peaking factor allows for j l the fact that power production varies in an assembly.
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 4 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUE5iTION 1.03 (1.50)
DEFINE the following:
'c) Quadrant Power Tilt Ratio b)-Axial Flux Diffevence I
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'l$[ PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE '5 l THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW 4
1 II QUECTION 1.04 (1.00)
Using the figure below, PLOT the " Xenon Concentration versus i Tido" curve that is associated with the given " Percent Full Pod:r versus Time" curve - i.e.g'Show how Xe concentration wruld respond to the given power changes, i
=
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- 1. PRINCIPLES OF' NUCLEAR POWER PLANT OPERATION. PAGE 6 !
THERMODYNAMIC 9. HEAT TRANSFER AND FLUID FLQH-t I GUESTION 1.05 (1.50) !
TRUE or FALSE {
a) Xc-135 is removed from the reactor by either decay to Cs-135 or by absorption of a neutron to form Xe-136. I b) Xenon equilibrium is achieved when the removal rate is r equal to the production rate and xenon concentration becomes constant with time.
c) The equilibrium concentration of xenon, and therefore its reactivity, varies in a non-linear manner with power level.
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- 1. PRINCIPLES OF' NUCLEAR POWER PLANT OPERATION. PAGE 7 .
THERMODYNAMIC. HEAT TRANSFER AND FLUID FLOW !
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l QUEGTION 1.06 (1.00) I 1
. DEFINE Shutdown Margin (SDM) as noted in Farley Station's l Tcchnical Specifications. l l
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1.4 PRINCIPLES OF' NUCLEAR POWER PLANT OPERATION.- PAGE -0 THERMODYNAMICS. HEAT-TRANSFER AND FLUID FLOW l
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- 1. PRINCIPLES OF NLICIFAR P M R PLANT OPERATION. PAGE 9 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW GUESTION 1.07 (1.00)-
o
. SCENARIO:
A fire has caused the evacuation of the control room and operations have been shifted to the Hot Shutdown Panel (HSP). Reactor /RCPs/ turbine have been tripped.
In order to verify proper hot standby conditons - as per step 5.15, AOP-28.2, " Fire In The Control Room".-
the following plant parameters are noted at HSP "A :
_-----------------_____--______-------_____----- .I RCS Wide Range. Temp =
- 625 degrees F
[N1B331TIO413 (T-HOT)]
Pressurizer Pressure =
- 1875 PSIG (PI-04442)
Pressurizer Level =
- 35 percent (LI-04592)
AVG S/G Wide Range Lvis =
- 68 percent (LI-477A, 487A and 497A))
RCS Wide Range Temp =
- 545 degrees F
[N18331TIO413 (T-COLD)]
T-Cold is near sat temperature for S/G pressure & j S/G' pressure is being controlled by S/G safeties. '
Thermocouple Temperatures Unavailable - System l Malfunction. j Using the above parameters and standard steam tables, '
CHOOSE / SELECT-the most accurate description of conditions thnt would be present in the RCS:
a) Saturated (slightly "superheated") core conditions exist in the RCS - Possible bubble formation in vessel - natural circulation conditions NOT met.
b) Subcooled core condition MAY be present in the RCS
- Subcooling Margin (SCM) is less than 10 degrees F - natural circulations conditions questionable. I c) Subcooled core conditions exist in the RCS - SCM ,
- l. As greater than 10 degrees F but less than 50 !
degrees F - natural circulation conditions met. !
d) Subcooled core conditions exist in the RCS - SCM greater than 50 degrees F - NC conditions met.
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r 1.- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 10 1 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.08 (1.50) i
- LIST FIVE (5) indications of natural circulation.
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1.. P3tNCIPLES'OF NUCLEAR POWER PLANT' OPERATION. .,
PAGE 11 l THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW
- f QUESTION i.09 (1.50)
NOTEn For the following questions - DISREGARD ANY ENERGY' )
INPUT FROM RX COOLANT SYSTEM (RCS) PUMPS, CHARGING FLOW, AND RCP SEAL IN-LEAKAGE AND ANY ENERGY LOSSES FROM LETDOWN AND PRIMARY SYSTEM PIPING.
Giv n that the rated thermal power-of each Farley Unit core 10 2652 Mw(t) and that the specific heat is approximately 1.39 BTU /lba - Degree F SHOW ALL CALCULATIONS - NOTE ALL ASSUMPTIONS c) During normal operations, what is the flow rate (in Ibm /hr) required to keep the temperature across the !
reactor vessel at 67 degrees F7 l
b) During normal operations, what is the approximate enthalpy rise / change (in Delta h) across the reactor core? l (NOTE: Include Units) c) [ CHOOSE THE MOST APPROPRIATE ANSWER] "With a reduction
-in RCS flow and with rated thermal power remaining at 2652 Mw(t) the enthalpy rise / change across the reactor will..(INCREASE, DECREASE, REMAIN THE SAME)."
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- 1. PRINCIPLksOFNUCLEARPOWERPLANTOPERATION. PAGE 12 THERHODYNAMICB. HEAT TRANSFER AND FLUID FLOW 4
GUESTION' 1.10 (2.00)
The following data was given to you for possible adjustments to the power range instruments.
- - - - - - - - - - - . - - - - - - _ - - - - _ = - - - = -- ------------- =-
I. Present Power Range Instrumentation Readings: 100 percent Control Bank Positions: Gp C = 228 steps Gp D = 220 steps II. Present MW(e) (gross): 860 = 97.8 percent III. Feedwater Temperatures 427.5 Degress F Feedwater Flow Steam Pressure Loop [(E +6) lbm/hr] (PSIG)
A 3.772 786 ,
i B 3.791 783 C 3.784 785 Using the steam tables provided, CALCULATE / DETERMINE the accociated percent THERMAL power output of the RCS.. !
NOTE: ASSUME ALL REACTOR THERMAL OUTPUT IS PLACED INTO THE SECONDARY - DISREGARD ANY ENERGY INPUT FROM REACTOR COOLANT SYSTEM PUMPS, CHARGING FLOW, AND RCP SEAL j IN-LEAKAGE AND ANY ENERGY LOSSES FROM LETDOWN AND !'
PRIMARY SYSTEM PIPING.
- SHOW ANY REQUIRED CALCULATIONS - NOTE ALL ASSUMPTIONS l
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- 1. PRINCIPLES OF-NUCLEAR POWER PLANT OPERATION. PAGE 13 l THERMODYNAMICS. HEAT TR @ FER AND FLUID FLQM l l
l QUESTION 1.11 (1.50)
SCENARIO: l
- During a recent out, age, THREE (3) percent of the "A"
.S/G. tubes were' plugged.
l "B" and "C" S/G's were not inspected nor plugged.
TRUE or FALSE Dua to the reduction in the number of tubes available in the )
"A" S/G: l l
c) The total heat transfer RATE of the "A" S/G will be .l 1ess than that of thts "B" or "C" S/G's.
1 b) Available (total) plant power output (MWe) is reduced j because of the "A" S/G output pressure reduction and the subsequent effects to Psat versus Tsat.
c) The Heat Transfer Coefficient ("U") will increase in the "A" S/G and decrease in the other two S/G's.
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 14 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW l
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QUESTIDN 1.12 (1.50)
Choore the responses.("words") which correctly complete the fellowing sentences:
1 c) One way in which an operator can minimize the possiblity I of creating a " water hammer" is to start a pump with it's discharge valve (OPEN, CLOSED).
b) By controlling (PUMP SPEED, DISCH. VALVE POSITION) we !
control volumetric flow rate out of a positive displacement pump.
c) Ideally, when TWO (2) pumps are placed in a (PARALLEL, SERIES) configuration, maximum flow capacity produced is no greater than the capacity of DNE (1) pump. ..
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j"6bdh;;jfNhbh,2fhhdpNI*M40$v T's UiN' # '
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-1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 15 THERMODYNAMICS. HEAT TRANSFER AND __ FLUID FLOld
' QUESTION 1.13 (1.00)
Using the single " Pump Head (Hd) Versus Flow Rate (V)" graph provided below, SKETCH / DRAW:
c)' A " set" of TWO (2) Emergency Core Cooling System (ECCS) pump curves - ONE (1) High Head and DNE (1) Low Head -
SHOWING their relative position to one another.-
b) A " set" of-TWO.(2) " system characteristic curves" depicting:
- 1) ONE (1) " system curve" associated with initial reactor coolant system (RCS) pressure AND;
- 2) ONE (1) other " system curve" which would be present during a major reactor coolant system (RCS) LOCA.
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L.' PRINC N 'PAGE.16 THERMODYNAMICS. KAT TRANSFER AND FLUID FLOW
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l GUESTION 1.14 (1.50) l TRUE or FALSE a) Since pressure stresses are higher on the RCS core :
vessel' inner wall during cooldown, restrictions on !
cooldown are more limiting than heatup restrictions. I I
b) Due to the greater. risk of brittle fracture, rather I I
than cyclic f.atigue, RCS core vessel " tech spec" heatup/cooldown limits are more restrictive than PZR pressure vessel limitations.
c) The relief capacity of a single PZR code safety
(*345,000 lbm/tr) is adequate to relieve ANY overpressure condition (" potential PTS problem") which could occur during shutdown (MODE FOUR (4) or MODE (5). ;
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- 1. PRINCIPLES OF NiflFAR POWER PLANT' OPERATION _._ PAGE 17 -
' THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW I
GUESTION 1.15- (1.00)
When synchronizing the generater to the grid - in accordance with,' FNP-1-SOP-28.1, " Turbine Generator Operation" - the op rator'is directed to regulate turbine speed to slowly rotate the synchroscope in the fast (clockwise) direction.
CHOOSE / SELECT from the following parameter differences, the ;
I TWO (2) parameters that the synchrescope is indicatings.
a) Current and voltage differences j b) Voltage and frequency differences c) Phase and resistance differences d) Resistance and current differences e) Frequency and phase differences
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.1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION 'PA3E 18-TE RMODYNAMICS, MAT TRANSFER AND FLUID FLOW QUESTION 1.16 (1.00) 1 Which ONE-(1) of the following statemento in CORRECT concorning the paralleling ~of electrical systems?
a) If the incoming machine is at synchronous speed but out of phase with the running bus when true i breaker is closed, heavy currents will flow to f Wither accelerate or retard the incoming machine. j l
b) If resistances are not matched when the {
synchronizing switch is closed, heavy currents l will flow - this tends to speed up the incoming machine to synchronous speed.
c) If voltages are not matched when the synchronizing switch is closed, there will be VAR flow from the i Iower voltage source to the higher'one.
d) If the incoming machine is in phase but slightly faster than synchronous speed when paralleled, the system will tend to speed up to synchronous speed.
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- i. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 19 j THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW l
QUESTION. 1.17 (2.50) !
SCENARIO:
The plant is operating at 20 percent power, turbine in AUTO (IMP IN), when loop "A" reactor coolant pump trips Accuaing no reactor trip, no operator action and rod control in MANUAL, indicate whether the following parameters will be HIGHER, LOWER, or the SAME at the end of the transient conp:1 red to their -initial values.
c) Loop "C" RCS flow b) Loop "B" S/G steam pressure c) Loop "B" RCS T-HOT d) Loop "A" RCS T-COLD o) Nuclear Power 1
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1.- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 20 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW GUESTION 1.18 (1.00)
An'oteam goes through a throttling process indicate whether the following parameters will INCREASE, DECREASE cir REMAIll THE SAME.
c) Pressure b) Enthalpy )
- c) Temperature d) Entropy
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT O,PERATION. PAGE 21 THERMODYNAMICS. HEAT TRANSFER AND F1UID FLOW '
QUESTION 1.19 (2.00)
"Dircct" surveillance an the," Heat Flux Hot Channel Fhetor" (F(QZ)] and the "Enthalpy Rise flot Channel Factor" [F(Delta b)]
cro only perforrad on a periodic basis .
LIST the FOUR (4) " specific" operational limitations - that !
cro found in FNP's Technical Specifications - which:, if !
" cot", assure that F(Delta h)(and F(QZ)'are within limits b2 tween " direct" surveillancus.
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i l 1. PRINCIFLES OF NUCLEAR' POWER PLANT OPERATION. PAGE 22 !
IBERMQDYlMMICS, HEAT _ TRANSFER AND FLUID FLOW l
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L QUESTION i.20 (2.00)'
Fer cach of the following events, state whether the differential rod worth of an individual control rod will l INCREASE, DECREASE or REMAIN THE SAME: l NOTE: Consider each case separately o) An adjacent rod is inserted to the'same height.
b) Moderator temperature is increased.
c) Baron concentration is decreased.
d) Fuel adjacent to the control rod is depleted.
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l 1. ' PRINCIPLES OF' NUCLEAR POWER PLANT OPERATION, '
PAGE 23 -l.
IbERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW.
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' QUESTION 1.21 (0.50) j i
TRUE or FALSE Incrcasing condensate depression (subcooling) will cause
.COTH a decrease in plant efficiency AND an increase in cond:nsate (hotwell) pump available NPSH.
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' 71 Pf.hMT-DESIGN INCLUDI'@ SAFETY AND EMERGENCY SYSTEMS PAGE 24 l-(
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-QUESTION 2.01 (1.00)
Given that Waste and Recycle System Components and Sample cGyct;m Heat Exchangers are serviced by the Component Cooling j W3 tar dC.CW) System, LIST EIGHT (B) "other" loads (both ESS l cnd' Secondary) which-are serviced by CCW.
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l2 . PLANT' DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 25 l
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QUESTION 2.02 (1.00)
The Component Cooling Water (CCW) System Surge Tank is divided into TWO (2) sections by a metal partition that extends from the bottom of the tank to about three-fourths of the height of the tank.
What is the purpose of this tank separation? (1.0) 6
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 23 l
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GUESTION 2.03 (2.00) l Unit ONE (i) has experienced a LOCA resulting in a Safety j InJcchion (SI) and all related equipment has functioned j properly.
For the following Service Water System (SWS) loads, note if i the designed SWS flow to the load is " GREATER THAN", "LESS l I
THAN" or of the "SAME" magnitude as it was before the LOCA.
c) Component Cooling Water Heat Exchangers (0.2) ;
.j b) Containment Coolers (0.2) l c) Control Room Air Coolers (0.2) j l
d) RHR (LHSI) Pump Room Coolers (0.2) o) Charging (HHSI) Puep Room Coolers (0.2) f) Auxiliary Feedwater Pump Room Coolers (0.2) I g) Bettery Charging Room Coolers (0.2) h) RCP Motor Air Cooler (0.2)
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- 1) Turbine Building Heat Exchangers (0.2) j) Diesel Generators (0.2) l l
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 27 i
GUESTION 2.04 (1.00)
Stato the purpose of the mini-flow recirculation lines inctalled between the downstream sides of Unit DNE (1) -
r RHR heat exchangers and the suctions of the RHR pumps. (1.0) 1' 1
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- 2. PLANT DESIGN INCLUDING SAFETY.AND EMERGENCY SYSTEMS PAGE 29 QUESTION 2.05' (3.00)
Concorning the Containment Spray (CS) Systems c) State TWO (2) purposes of the Sodium Hydroxide (NaOH) which is added to the spray solution. ,
b) What CS finw path (design) provision is available for running a CS Pump performance test?
c) State TWO (2) suction sources for each spray pump.
(NOTE: DO NOT include the eductors.)
b
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 29 i
I l QUESTION 2.06 (2.00)
Sp;cify location - in terms of" Loop A", " Loop B" or " Loop C" AND " Hot Leg", " Cold Leg" or " Intermediate Leg" - for'the l fallowing Reactor' Coolant System _(RCS) penetrations.
c) Normal Letdown Line b) Pressurizer Surge Line c) Excess Letdown Line
-d) Normal Charging Lines o) Alternate Charging Lines ,-
- 1) Pressurizer Spray Lines [ NOTE: TWO (2) required]
g) Tygon.Home Connection [ NOTE: For Reactor Vessel Level '
Measurement During Refueling]
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTfdiR PAGE 30 QUESTION 2.07 (1.50)
Conecrning the Auxiliary Feedwater (AFW) Systems
! c) What percentage of the decay heat removal capacity is the AFW system capable of, if DNE (i) motor driven pump AND the turbine driven AFW pump is out of service?
b) LIST TWO (2) sources of air which can supply the TDAFW Pump air-operated isolation valves * (3235A and 32359) cmergency air reservoirs.
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- 2. PLANT' DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 31 QUESTION 2.08 (1.00)
Whnt is the SPECIFIC purpose for each of the following !
limitations on ECCS Accumulator parameters 7:
c) High Water Level (Max Vol of 7700 gals /58.4 percent)(0.5) )
b) High Nitrogen Pressure (Max Pressure of 649 PSIG) (0.5) r I
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l 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 32 l
I t-QUESTIDN 2.09 (1.00)
Wh t set of signals below is sent to the Reactor Protection Eyctcm to indicate a Turbine Trip?
o) Throttle valves closed and Auto Stop 011 pressure low b) Throttle valvem closed and EHC pressure low c) Gov.trnor valves closed and Auto Stop 011 pressure low d) Governor valves closed and EHC pressure low
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 33 I i
i, QUESTION 2.10 (1.00) 1 By d sign, the maximum actual capacity of any single Main Stocm Atmospheric Relief Valve, at a steam pressure of 1085 PSIG, will not exceed 890,000 lb/hr.
Whst is the reason for this limitation?
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2 .' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 34
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i GUESTION 2.11 (1.00) i Conccrning the Unit DNE (1) Waste Gas System (GWD):
c) How many " shutdown" gas decay tanks are there? (0.5) b) Why is the waste gas distributed among all normal l cervice gas decay tanks instead of filling one tank at a time? (0.5) i l
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 35 i
QUESTION 2.12 ! (1.00) !
-Which DNE (1)- c,f the following completes this statement:
"During Unit UNE (1) plant cooldown, FLOW from the RHR to the RCS is designed to be ...:
i a) ... constant (around 3000 gpm)-AND this total flow is maintained by controlling RHR heat exchanger (HX) flow bypass valve (FCV-605A or FCV-605B)".
b) ... constant (around 3000 gpm) AND this total flow is maintained by controlling RHR heat exchanger (HX) discharge valve (FCV-603A or FCV-603B)".
c) ... varied (to a flow that meets desired cooldown rate) AND this flow is maintained by controlling RHR HX bypass valve (FCV-605A or FCV-605B)".
d) ... varied (to a flow that meets desired cooldown rate) AND this flow is maintained by controlling RHR HX discharge valve (FCV-603A or FCV-603B)". ;
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'2. PLANT DESIGN INCLUDING SAFETY ANDJMERGENCY SYSTEMS PAGE 36 l
QUESTION 2.13 (1.50) .
l Concerning ttie Unit ONE (i) Post-LOCA Atmospheric Control l System (and subsystems):
c) State the purpose of the Post-Accident VentinD Subsystem.
( 0. 5 ) '
b) State the purpose of the Post-LOCA Air Mixing Subsystem.
(0.5) c) TRUE or. FALSE The Hydrogen Recombiners use natural convection flow cs part of the recombiner operation. (0.5) j 1
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'2.- PLANT DESIGN ~ INCLUDING BAFETY AND EMERGENCY BYSTEMS PAGE 37 ;
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= QUESTION 2.14 (1.00)
-Conecrning Unit DNE (1) 120V Vital AC Instrumentation powers Which DNE (1) of the following electrical flowpaths most -
eccurately describes how power is NORMALLY supplied to a typical 120V AC instrumentation panel (1A, 1B, 1C or 1D)?
n) 600 VAC from 600 emergwncy supply, inverted to *120 VDC, rectified to *120 VAC, then supplied to the panel, b) 600.VAC from 600 emergency supply,-rectified to *125 VDC, inverted to *120 VAC, then supplied to the panel.
c) 208 VAC from 600 emergency supply, transformed to *120 VAC, then supplied to *125 VDC backup and to the panel.
d) 125 VDC from auxiliary building distribution, rectified to *120 VAC, then supplied to the panel.
o) 125 VDC from auxiliary building distribution, inverted to *120 VAC, then supplied to the panel.
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 33 15 0 QUESTION 2.15 M i LIST the FOUR (4) sourcas of normal / emergency makeup water ;
to the Condensate Storage Tanks. l
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 39 i
5 I QUESTION 2.16 (2.00)
Sinco the Reactor Coolant Pumps (RCP's) are not designed for '
frcquent " start-stop" operations, what starting duty requirements must be observed?
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMji PAGE 40 l
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' GUESTION 2.17 (2.00)
Concerning the Chemical and Volume Control (CVCS) System:
c) Fill in the matrix below . State whether the below valves are OPEN or CLOSED - during Reactor Coolant -i System makeup - for each of the indicated modes of
. operation.
l NOTE: Figure 2-1 is attached for reference. l FCV-113A FCV-113B FCV-114A FCV-114B >-
" DILUTE" .j i
" BORATE"
" ALTERNATE" DILUTE 1
b) IF the " Reactor Makeup Select Switch" were in the "AUT0" position, what parameter would control'.both the starting cnd stopping of automatic CVCS makeup? i I
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-D<3- ~ E - .
Qx17 E i5 n >c+ I i -C<30 ! " ~) (i ! !) y .
5 0 '
s -l .
g '
g I- . i fr \!E: 5 l l~ Eg m
- E -::-
80-@ -l
~
- ! h::-
IZ E E _ =
s Q' l e
T k(-<1 ' '
5~UhD Oni I
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REACTOR MAKEUP AND CHEMICAL ADDITION SYSTEM l
FIGURE 2-1
-2.- PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 41' l GUESTION 2.18 (3.00)
Conccrning Engineered Safety Features (ESF) systems:
a) LIST EIGHT (B) of TEN (10) ESF systems.
b) LIST FOUR (4) ESF " SUPPORT" systems. j l
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY BYSTEMS PAGE 42 I I
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. QUESTION 2.19 -(2.00)
In FNP's Loss of Coolant (LOCA) " accident" analysis it has b;cn'noted that a "large-break" LOCA consists of FOUR (4) cherceteristic stages. j LIST AND DESCRIBE _the FOUR (4) stages. Include in your entw:r, the " time frames" associated with each of these
- "rccognized" stages.
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L3. INSTRUMENTS AND CONTROLS PAGE 43' 1 00
' QUESTION 3.01 m)
I c) What PROTECTIVE function can~'be manually blocked by the operator when 2/3 (TWO out of THREE) pressurizer I
pressure channels fall below 2000 psig (P-11)7 a) LIST the TWO (2)~ PROTECTIVE functions which are " auto" actuated when 2/3 (TWO out of THREE) T-AVG instruments fall below 543 degrees F AND high steam flow conditions exist (P-12).
c) As RCS T-4VG falls below the "Lo-Lo T-AVG" setpoint, all open steam dump valves shut and any further system operation is " blocked". How is this specific form of CONTROL restored if further plant cooldown is desired?
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- $ INSTRUMENTS'AND CONTROLS PAGE 44~
QUESTION 3.02 (1.50)-
Concorning the Compressed Air Systems a) LIST FOUR (4) of the FIVE-(5) signals which will automatically trip (shutdown) the "2D" Air Compressor.
b) FILL IN THE BLANKS The number of " normal" 2D Air Compressor starts should be limited to a maximum of _ starts per hour.with an interval of minute (s) between starts.
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3.a__lNEIBWEiTS_8HD_G2iTEDLR PAGE 45 l
!: QUESTION 3.03 (1.00)
Which ONE-(1) of the following statements most correctly l l doccribes the effect.of " overcompensating"-Intermediate Ranga Detector (IRD) N-357 NOTE: Assume all other Nuclear Detectors are functioning properly.
c) The IRD N-35 will indicate a HIGHER power level than the actual power level AND the Source Range Detectors (SRDs) WILL automatically energize above their reset setpoint (10exp-10 amps) during a reactor shutdown. .
b) The IRD.N-35 will indicate a HIGHER power Invel than the actual power level'HOWEVER the SRDs WILL NOT automatically energize above their reset setpoint i (10exp-10 amps) during a reactor shutdown.
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c)-The IRD N-35 will indicate a LOWER power level than j the actual power level AND the SRDs WILL automatically j energize above their reset setpoint (10exp-10 amps) I during a reactor shutdown.
l d) The IRD N-35 will indicate a LOWER power level than the actual power level HOWEVER the SRDs WILL NOT i automatically energize above their reset setpoint
{10exp-10 amps) during a reactor shutdown.
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- 3. INSTRUMENTS AND CONTROLS PAGE 46 l
i QUESTION 3.04 (3.00)
Concerning Nuclear Instrumentation - Power Range:
LIST, 1) all PERMISSIVES, 2) associated LOGICS and 3) a BRIEF DESCRIPTION of what is-accomplished by each permissive for Power Range Instruments (N41 through N44).
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[' 3. INSTRUMENTS AND CONTROLS PAGE.47 l
l QUESTION 3.05 (2.00)
Conccrning the Full-Length Rod Control Systems o) LIST FOUR (4) of FIVE (5) conditions which can activate
- a Power Cabinet " ROD CONTROL URGENT FAILURE" alarra (annunciator).
b) LIST TWO (2) of THREE (3) " automatic" actions which occur upon activation of the above URGENT FAILURE alarm.
c) TRUE or FALSE Upon activation of the " ROD CONTROL URGENT FAILURE" olarm, rod motion is still available in the other "non-affected" power cabinets.
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- 3. INSTRUMENTS AND CONTROLS PAGE 48 GUESTION 3.06 (2.00)
Conecrning Steam Generator (S/G) Water Level Control l c) What is the purpose of the turbine impulse pressure-cignal used in the S/G Water Level Control System?
b)~ Considering only the Feedwater Regulating Valve Control portion'the S/G Water Level Control System:
- 1) Indicate whether feedwater flow would initially ) '
INCREASE, DECREASE or REMAIN THE SAME, AND...,
- 2) Driefly explain your choice for the following ;
1 situation:
The controlling S/G pressure transmitter fails HIGH during 50 percent power operation.
NOTE: Assume turbine feed pump speed control is in manual.
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- 3. INSTRUMENTS AND CONTROLS PAGE 49 i
-QUESTION 3.07 (1.50)
Concarning Steam Generator (S/G) Water Level Control:
LIST THREE (3) variable plant parameters which provide an input to the feedwater pump speed control circuitry.
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MEMLE PAGE 50'
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GUESTION. 3.08 (2.00)
Unit ONE (1) is operating at 100 percent rated' thermal pnwar. The' rod control system is in manual control - all other systems are in automatic control and operating normally.
For cach of the following instrument failures, STATE how proccurizer (PZR) level is affected, AND WHY PZR level recponds in the manner present.
NOTE: Assume NO operator action is taken AND limit your response'to include the period up to when the plant is stable or the reactor. trips - whichever occurs first.
i c) PZR Level Channel DNE (1) - the controlling channel - ;
fails HIGH.
.b) Loop DNE (1) RC's T-AVE fails HIGH.
c) Turbine Impulse Pressure Transmitter fails to 85 percent power position.
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(- 3. INSTRUMENTS AND CONTROLS PAGE Si l
( QUESTION 3.09 (2.50)
Conecrning the Emergency Diesel Generators (DGs) -
c) An autostart signal for a diesel will energize its cmergency start (ES) relay provided TWO (2) conditions j ore met with the diesel's " mode-related" switches. 1 What are these " switch" conditions?
b) LIST SIX (6) of SEVEN (7) "Non-essential" protection i chutdowns which are blocked when the ES relay is cnergized.
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- 3. INSTRUMENTS'AND CONTROLS PAGE 52 i
OUESTION 3.10 (1.00)
Concerning the Emergency Diesel Generators (DGs):
What arar the positions available on a swing diesel's " UNIT SELECTOR SWITCH" AND what " protective condition" has the oporator placed the swing diesel in for these positions?
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- 3. INSTRUMENTS AND CONTB ER PAGE 53 QUESTION 3.11 (2.50)
Conccrning the. Steam Dump System:
In comparing the " Loss-of-Load" controller with the " Turbine Trip" controller, what differences exist (IF ANY) between occh controller in regard to the following7:
c) Point at which the " Bank DNE (3)" steam dump valves l ctart to open. l l
'b) "HIGH ONE (1)" trip open " quick" steam dump valve l response (Banks 1 and 2) - delta T setpoint.
c) "HIGH TWO (2)" tri pen " quick" steam dump valve response (Banks-1 and - delta T setpoint.
emus Y-._, s 4 6c.A %
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'3. INSTRUMENTS AND CONTROLS PAGE 54 l
l QUESTION 3.12 .(3.00)
Conecrning the Steam Dump System Control Interlocks:
l l LIST, 1) all CONTROL ("C") INTERLOCKS, 2) associated LOGICS / CONDITIONS (if any) and 3) a BRIEF DESCRIPTION of' whnt is accomplished by aach interlock.
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I 3.' INSTRUMENTS AND CONTROLS PAGE 55 )
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. QUESTION 3.13 (1.50) u, i C:necrning the Fuel Handling Equipment Systems
-Thero are THREE (3) " main" controls for the fuel manipulator I '
crcno - NAME them.
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- 3. - INSTRUMENTS AND CONTROLS PAGE 56-
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QUESTION 3.14 (1.50) i
.Concorning the Unit ONE (1) Fuel Handling Equipment Systems LIST THREE (3) of-FIVE (5) conditions (" interlocks") which cust be met before the Fuel Transfer System Conveyor can be covcd.
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- 3. INSTRUMENTS AND CONTROLS PAGE 57 >
QUESTION 3.15 (1.50)
Concarning the Radiation Monitoring System - Specific' ally th2 Spent Fuel Pool Exhaust Flow Gas Monitors R-25A and B Whnt "automs. tic" actions occur when gaseous activity in the Sp:nt Fuel Handling Building exhaust line reaches the high Icvol setpoint? i 4
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(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
l 3 .- INSTRUMENTS AND CONTROLS PAGE 53 l l
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QUESTION' 3.16 (1.00)
Conccrning the Emergency Core Cooling System Cold Leg Accumulators:
Ono dcsign feature of the motor-operated Accumulator Icoletion Valves-(8808 A, B and C), is that, if energized, during a plant startup, they can be operated-(opened) from tha Mnin Control Board (MCB) upon reaching 1000 PSIG.
DESCRIBE THREE (3) "other" design features / interlocks o cocociated with operating these valves: I 1
(***** END OF CATEBORY 03 *****)
4.' PROCEDURES - NORMAL 4 @ NORMAL. EMERGENCY AND PAGE 59 1 RADIOLOGICAL CONTROL ,
' QUESTION 4.01 (1.50)
'A portion of the bases for Technical Specification 3.4.7.2, "Opsrotional Leakage" states that:
"The total steam generator' tube leakaga limit of DNE (i) RPM for all steam generators ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break."
c) What are the dosage limits'of both the Low Population ,
Zone (LPZ) and the Exclusion Area?'(10 CFR 100 Limits) b) H:w do event " time frames" - associated with the 10 CFR Port 100 limits - DIFFER between residents of the " Low Population Zone (LPZ)" and the individuals located on the bountry of the " Exclusion Area"?
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- 4. PROCEDURES ' - NORMAL. ABNORMAL. EMERGENCY AND PAGE 60 BSDIOLOGICAL CONTROL l-s1UESTION 4.02 (2.00)
Concerning Engineerd Safety' Features (ESF) System Designs i In order for the plant to ensure that ESF functions will be ccrried out, FNP design must assume that a " single failure" occurs.
DESCRIBE / EXPLAIN what is meant by " single failure criteria"?
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.. E
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-4. PROCEDURES - NORMAL.. ABNORMAL. EMERGENCY AND PAGE 61 .
BentotaalG8L_GoNIBQL j l
1
'GUESTION 4.03 (1.00) i Th3 only people allowed to manipulate any control that j dircctly affects reactivity.or power level are those who i currcntly hold an NRC RO or SRO license AND ...: j a) ..by those in training for an RO/SRO license IF3 ;
i) . prior SRO approval has been received
- 2) individual is under direction of a licensed RO/SRO
- 3) individual is in the presence of a licensed RO/SRO j l
l b) ..by those in training for an RO/SRO license ifs i) prior SRO approval has been received l
- 2) individual is under direction of a licensed RO/SRO !
I c) ..by those.in "non-licensed technical" positions IFg i) prior SRO approval has been received !
- 2) individual is under direction of a licensed RO/SRO
- 3) individual is in the presence of a licensed RO/SRO i
- 4) the job requires direct support from within the individual's field of expertise d) ..by those in "non-licensed technical" positions IF3 ;
i) prior SRO approval has been received !
- 2) the job requires direct support from within the individual's field.of expertise i
(***** CATEGORY 04 CONTINUED ON NEX'i PAGE *****)
1 j
t; I 4 '. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 62 j RADIOLOGICAL CONTROL '
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GUESTION 4.04 -(1.00)
After transferring previously prepared (" batched") boric ,
ccid-from the BA Batch Tank (BABT) to storage tank "1B", the
.opsrator (according to procedure FNP-1-SOP-2.6) musta
- 1) Verify 1A BAT Boric Acid Concentration is 4.0 to '
4.4 w/o as indicated by latest chemistry analysis.
- 2) Place 1A and 1B. Boric Acid transfer pump control switchen in STOP and return to AUTO.
- 3) Shut batching tank supply to Boric Acid pumps ,
1-CVC-V-8310.(QiE2iV236). l
'The chift supervisor has determined that it.is desirable to parform steps TWO (2) and THREE (3) prior to the " chemistry" verification called for in step ONE (1). You agree with 4 the decision.
. Aanuming that the ' plant is in " normal" MODE ONE (1) op3rction and the system is functioning properly - what .
cdninistrative guidelines would allow these procedural steps !
to be performed "out-of-sequence"?
- STATE ANY ASSUMPTIONS TAKEN TO SUPPORT YOUR ANSWER -
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~ 4 '.' PROCEDURES - NORMAL.' ABNORMAL. EMERGENCY AND PAGE 63 j RADIOLOGICAL CONTROL l
l' I
QUESTION 4.05 (2.00) i l
\
l LIST the FOUR (4) immediate operator actions of the i
l ECCrgency Boration procedure FNP-1-AOP-27.0. j l
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PAGE 64.
~
'4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND BBDlQLDEIC81._GQNIfM2L QUESTION 4.06- (2.00)
Using the attached drawing, Figure 4-1, LABEL / IDENTIFY the u following " Emergency Plan" areas:
a) Operations-Support Center (Operations /HP/ Chemistry /Env.)
b) Technical Support Center c) Alternate Technical Support Center l
d) Central Alarm Station o) Controls Area i
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(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
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'4. PROCEDURES.- NORMAL. ABNORMAL, EMERGENCY AND PAGE 65
, RADIOLOGICAL CONTROL
..,~
- j. QUESTION 4.07 (1.00)
SCENARIO:
l During-Mode ONE(i) operation, a control rod drive l mechanism (CRDM) is ejected from the NC-vessel head, becomen an internal missile hazard and breeches the containment structure.
A " Loss of Containment Integrity" means that ONE (i) of the l Tcchnical Specif.ication items which denotes " Containment Intcgrity" is " violated".
Which ONE (i) of these items is " violated" in the above cccncrio?
- i,
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)
4.- PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND' PAGE 64- I t
ftADIOLOGICAL CONTROL i
QUESTION 4.08 (1.50)
~TRUE or FALSE !
a)- A-" continuous" fire watch must be able to continuously detect a fire in ANY part of the af fected area. l 4
b) An " hourly" fire watch may perform-other plant duties-concurrent with the actual fire watch tasks.
c) A " continuous" fire watch will ALWAYS require backup fire suppression equipmet.
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I l i j
% )e 1/ PROCEDURES - NORMAL. ABNORMAL. EPERGENCY AND PAGE 67 RADIOLOGICAL CONTROL l
l QUESTION 4.09 -+2.GG)
SCENARIO:
A fire exists in the Control Room (around the center desk section) AND the Shift Supervisor decides that control of equipment from control is in jeopardy.
LIST FOUR (4) of FIVE (5) actions which should be performed 1
- in an expeditious manner and in accordance with the " Fire In The Control Room" procedure-(ADP-28.2) - both PRIOR TO end IMMEDIATELY AFTER the control room evacuation and BEFORE coccmblage of personnel at the Hot Shutdown Panel.
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1
- 4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY ANQ PAGE 6")
RADI 6 i
GUESTION. 4.10 (2.00)
Conccrning Logging Requirements for Reactor Power Changes (FNP-84-0220) - Operations Memorandum 84-03:
Dua to the reporting requirements of NERC-GADS and the
" Performance Monitoring Report" (FNP-0-AWP-34), a means was catchlished for reporting reactor power reductions.
c) What " magnitude" of power reduction (" change") must be reported to FNP's Licensing Group?
-b) What " units" are used when making this report? - i.e.:
NI Indication (percent P/R change)?
Percent of Gross MW Elec. Generated?
Percent Rated Thermal Power (RTP)?
Gross MW minus Reactive MW Generated?
l c) LIST the FOUR (4) " pieces" of information that should be cntered into the Reactor Operator's logbook.
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- 4. ' PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 69, -
RADIOLOGICAL CONTROL I l
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- QUESTION 4.11 (1.50)
Concerning Surveillance Scheduling: ]
l NOTE: Refer to Figures 4-2 and 4-3 for the following .
questions. j a) What is the periodicity of surveillance listed on the "B-Schedule" (Figure 4-3)?
b) How are the "DUE DATES" - as listed on the "A-Schedule" (Figure 4-2) - determined?
c) How are the " GRACE PERIOD END DATES" - as listed on the "A-Schedule" (Figure 4-2) - determined?
1
(***** CATEBORY 04 CONTINUED ON NEXT PAGE *****) {
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- 4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 70 1 RADIDLDGICAL CONTROL 1 DUESTION 4.12 (1.50)
Concerning FNP's Access Control Requirements:
Controlled Areas of FNP include internal areas of the l Auxiliary'Bu11 ding that have been locked and alarmed in )
cccordance with Chapter 11 and Table 5.12 of the FNP S:curity Plan-for the implementation of engineering changes and administrative upgrading to improve the seaurity of FNP. j LIST SIX (6) of these areas.
' NOTE: Listing of "1A Blank Equipment Room" and "iB Blank Equipment Room" will count as a single " Blank Equipment Room" NOT as TWO (2) separate areas.
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- 4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 71 RADIOLOGICAL CONTROL i
l THIS PAGE INTENTIONALLY BLANK i
4
-(88888 CATEGORY 04 CONTINUED ON NEXT PAGE $$888)
- 4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 72 RADIOLOGICAL CONTROL QUESTION 4.13- (2.00)
Uso Figures 4.4 thru 4.9 (Emergency Procedure CSF Drawings FNP-i-CSF-0.1 thru CSF-0.6) for the following question:
Unit ONE has just experienced an ATWT event, however, all rodo are now inserted, the turbine is tripped, and NO cdvsrse containment conditions are experienced.
SPDS is unavailable but from other. indications'you note that the following Critical Safety Function. conditions exist
.o PRESSURIZER LEVEL IS '63 PERCENT AND " STEADY" WITH NO NOTICEABLE FLUCTUATION IN LEVEL o CONTAINMENT SUMP LEVEL'IS *THREE (3) FT. AND RAD MONITORS INDICATE *120 MREM o " INTEGRITY" CSF DETERMINED TO BE " SAT" o NR LVL IN ALL S/G's LESS THAN FIVE PERCENT - TOTAL FDWTR FLOW TO ALL S/G'S '400 SPM - AVG PRESS *1077 PSIG o NC. SYSTEM SUBCOOLING IS *ZERO (0) DEGREES F AND FIFTH HOTTEST CORE EXIT TC AVG TEMP '715 DEGREES F y a POWER RANGE INSTRUMENTS INDICATE THREE (3) PERCENT INTERMEDIATE RANGE STARTUP RATE SLIGHTLY POSITIVE c) Which of the above conditions, require at least
" PROMPT" (" ORANGE") actions / responses?
b) If.the Power Range instruments suddenly indicated greater than FIVE (5) percent reactor power, the operator would:
- 1) Continue with functional restoration of the " loss of heat sink" condition THEN commence functional I
restoration =nf the " loss of subcriticality" condition.
- 2) Proceed with functional restoration of the " loss of subcriticality condition" THEN continue with restoration of the " loss of heat sink" condition.
- 3) Continue with functional restoration of the " loss of
, core cooling" condition THEN coeeence functional l restoration of the " loss of subcriticality" condition.
- 4) Proceed with functional restoration of the " loss of subcriticality condition" THEN continue with restoration of the " loss of core cooling" condition.
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'(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
FNP.1 CSF-0.1 SUBCRITICALITY REYlSION 3 GO TO FRP-S.1 .
' GO TO EEEEE FRP-S.1 NO POWER RNG. ~
- GO TO LESS THAN 5% gg FRP-s.2 YES g 4
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BOTHINT NO i RNG SUR-BOTHINT NO MORE -
RNG SUR ~ NEGATIVE
~
ZERO OR THAN 4.2 DPM YES NEGATIVE YES CSF SAT l -
BOTH NO SOURCE
- RNG ENERGlZED YES Go TO OO FRP S.2 4
4 BOTH NO SOURCE i RNG !
FIGURE 4-4 SUR ZERO OR '
l NEGATIVE YES
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......... 1 ;;:o NAR RNG NO
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PRESS IN NO ALL SG'S LESS THAN 1075 PSIG YES GO TO O 8 0 0#U FRP-H.5 NAR RNG NO '
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COLD LEG TEMPERATURE RP P 1 ALL RCS gg7g PRESS- NO 5MMMM pnp.p.3 !
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POINTS TO YES RIGHT OF ALL RCS NO ,, aoTo gly A CL TEMPS - FRP-P.2 L IN LAST e 60 MIN e GRTR THAN YES 2660 F ALL RCS -
NO CL TEMPS H
TEMP DECR - IN LAST '
IN ALL CL NO 60 MIN IN LAST - GRTR THAN YES 60 MIN 2960 F LESS THAN YES _
FIGURE 4-7 aoyo f FR P-P.1 NO ALL RCS CL TEMPS F GRTR THAN 2660F YES RCS PRESS NO LESSTHAN , go 7o f~ 450 PSIG ee pgp.P.2 YES j NO ALLHCS TEMPS -
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- 4. PROCEDURES - NOR2h . ABNORMAL. EMERGENCY AND PAGE 73 RADIOLOGICAL CONTROL
(
QUESTION' 4.14- (2.00)
Ona objective of Farley's Critical Safety Functions - and tho overall " purpose" of ESF design, in conjunction with the
. rocctor control / protection system design - is to ensure that FNP*o " multiple fission product barriers" remain intact and thnt FNP's policy of " defense in depth" is maintained.
I LIST these " multiple fission product barriers". l 1
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- 4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 74 RADIOLOGICAL CONTROL l
l
)
QUESTION 4.15 (1.50) l LIST SIX (6) " parts"/" sections" which makeup Farley's Tcchnical Specifications. l
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(________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _
- 4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 75 RADIOLOGICAL CONTROL QUESTION 4.16 (2.50)
Conecrning FNP's Radiation Work Permits:
c) Given that a Special Radiation Work Permit (SRWP) would b3 used for " specific tasks or series of tasks", LIST FOUR (4) of FIVE (5) specific situtations - as cutlined '
in FNP's Health Physics Manual, FNP-0-N-OOi - which would rcquire the issuance of a SRWP.
b) According to FNP's Health Physics Manual, FNP-0-M-OOi, when (under what condition / situation) would a " Routine" Radiation Work Permit be written?
c)-TRUE or FALSE In "special" cases, the presence of health physics p3rsonnel may be substituted for an RWP if prior cpproval has been given by the on-shift HP Foreman.
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~ 4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 76' RADIOLOGICAL CONTROL !
)
l DUESTION 4.17 (1.00)
Concerning " Authorized Deviations" from procedures and Tcchnical Specifications:
According to FNP-0-AP-6, " Procedure Adherence", when - under what set of circumstances / conditions / situations - are FNP Op; rations personnel allowed to depart from FNP procedures?
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- 4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 77 RADIOLOGICAL CONTROL QUESTION 4.18 (2.00)
I An operator has a documented Whole Body radiation exposure 1 of 750 mrom for the current quarter. The operator has been l
,cuthorized to exceed the normal allowable Whole Body limit 1 by the HP Manager.
H:w long can this individual stay in a 50 mrom/hr radiation croo without exceeding the HP Manager's authorized limit?
IMPORTANT!!! STATE ALL ASSUMPTIONS - SHOW CALCULATIONS i
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(************* END OF EXAMINATION *************)
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ANSWER KEY FARLEY RO EXAM REGION II 88/11/14 I
I l
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 78 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW J
- ANSWERS -- FARLEY' -88/11/14-REGION II ANSWER 1.01 (1.50) c) FALSE (0.5) b) FALSE (0.5) c) TRUE (0.5)
, REFERENCE l Fcrley LP OPS-31305A, Obj 22 & 233 LP DPS-31305B, Obj 3 l KAIR 3.3/3.4
- 192OOOK110 ....(KA*S) lL
- l. Sb ANSWER 1.02 (2 06) o) AXIAL (0.5) ti) DE5IGW (o3T N"""~
.c) RADIAL (0.5) d) LOCAL (0.5)
REFERENCE NUS, Module 4, " Plant Performance", Chapter 10, "PWR Performance, Section 10.2, Farley LP DPS-52102J AND W;otinghouseg " Nuclear Power Distribution & Core Control" Material KAIR 2.9/3.1 2.9/3.3 192OO5K112 193OO9K107 ...(KA'S) i
si . PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 79 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- FARLEY -88/11/14-REGION II ANSWER 1.03 (1.50) a) Quadrant Power Tilt Ratio - ratio of the maximum upper oxcore detector calibrated output to the average of the upper excore detector calibrated outputs, or. . . (0.45) the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs... (0.45)
...whichever is greator. (0.1) b) Axial Flux Difference - the difference in normalized flux oignals between the tcp and bottom halves of a two ecction excore neutron detector. (0.5)
NOTE: Exact Technical Specification wording IS NOT required for full credit, HOWEVER, a basic " idea"
(" understanding") of GPTR and AFD terminology should be expressed AND phraseology denoting
" standardization of signals" - i.e. g " calibrated output" and " normalization of output" - should be presented.
REFERENCE Farley Technical Specification Definitions 1.2 and 1.2.4 AND LP OPS-52201D, Obj 14 and 16 KAIR 3.0/3.3 2.9/3.3 3.3/3.7 3.2/3.6 192OO5K110 192OO5K113 015020K503 015020K504 ...(KA*S)
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 80' I THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW 1
ANSWERS -- FARLEY -88/11/14-REGION II 4
l ANSWER 1.04 (1.00) 66 1
1
- Place " master" drawing here - hf;II'1'A Look for the following:
a) Correct representation of Xe Conc change in response to ZERO (0) percent to 100 percent power change. (0.20) 1 i b) Correct representation of Xe Conc change in response to l 100 percent to ZERO (0) percent power " trip". (0.20) J c) Correct representation of Xe Conc change in response to ZERO (0) percent to 50 percent power change. (0.20) ,
1
'4 d) Correct representation of Xe Cone change in response to I 50 percent to 100 percent power change. (0.20) o) Correct representation of Xe Conc achieving equilibrium conditions while at 100 percent power. (0.20)
REFERENCE Farloy LP OPS-31304D, Obj 7 and 8; Glasstone & Sesonske;
" Nuclear Reactor Engineering"g Parts 5.90 through 5.101 i Lcmcrsh; " Nuclear Reactor Theory"; Pgs 469 through 474 KAIR 3.1/3.1 192006K111 ...(KA*S) l l
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 01 l THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- FARLEY -80/11/14-REGION II ANSWER 1.05 (1.50) c) TRUE (0.5) b) TRUE (0.5) c) TRUE (0.5)
REFERENCE Fcr1cy LP OPS-31304D, Obj 4, 5 and 63 Glasstone & Sesonskeg
" Nuclear Reactor Engineering"; Parts 5.90 through 5.94 3 Laanrsh; " Nuclear Reactor Theory"; Pgs 469 through 474 KAIR 3.0/3.1 192OO6K102 ...(KA*S)
ANSWER 1.06 (1.00)
"ths INSTANTANEOUS amount of reactivity (0.1) by which the reactor IS suberitical (0.1) or WOULD BE subcritical (0.1) from its PRESENT condition (0.1) cccuming ALL full-length rod cluster assemblies / RODS (0.1) cro FULLY INSERTED (0.1)
EXCEPT for (with the exception of) (0.1) the SINGLE rod cluster assembly / ROD (0.1) of HIGHEST reactivity WORTH (0.1) which is ASSUMED to be FULLY WITHDRAWN." (0.1)
NOTE: The words " ROD" or " RODS" are acceptable substitutes for the terms " ROD CLUSTER ASSEMBLIES" or " ASSEMBLY".
REFERENCE Foricy LP OPS-31305A, Obj 1 AND Unit Technical Specification (SDM) Definition 1.28 KAIR 3.2/3.6 192002K110 ...(KA*S) i I.
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. 'FAGE 82 INERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- FARLEY -88/11/14-RE8 ION II l l
l i
l ANSWER 1.07 (1.00) b) (1.0) l l
Tha following calculation IS NOT required for full credit )
/
Tect for 1875 PSIG (*1890 PSIA) = *627.9 degrees F l T-HOT Wide Range Temperature = 625.0 degrees F l l
Subcooling Margin = Tsat - T-HOT Temp l
Subcooling Margin =-*627.9 - 625.0 = *2.9 degrees F REFERENCE l
Fcr1cy LP OPS-30913, Obj 16; Steam Tables AND AOP-28.2, i "Firo In The Control Room"; Farley LP OPS-52202D, Obj ectives '
1 cnd 3 AND Qual Req'mt ADP's KAIR 3.3/3.4 193OO3K125 ...(KA*S) l l
l
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l I
,1. PRINCIPLES OF NUCLEAR POWER PLANT DPERATION. PAGE 83 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- FARLEY -88/11/14-REGION II 1
ANSWER 1.08 (1.50) i (any 6 al- 0 3 each) o) RCS Subcooling Monitor indicates "becoming more ;
subcooled" (0.3) j b) RCS T-hot' leg temperatures stable OR trending down(0.3) l c) Core exit thermocouple stable OR trending down (0.3) i d) S/G pressure stable OR trending down (0.3) o) RCS T-cold temperatures at saturation temperature for S/G pressures (0.3) f) RCS OT l- (o s F REFERENCE Fer1cy LP OPS-30913, Obj 16 AND Qual Req'mts ERP's/ ESP-0.2 KAIR 4.2/4.2 3.9/4.1 193008K122 193OOOK123 ...(KA*S)
__.__._________..-.____..m_ _ _
- 1. PRINCIPLES OF' NUCLEAR POWER PLANT OPERATION. PAGE C4 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW i
' ANSWERS -- FARLEY -B8/11/14-REGION II !
l ANSWER 1.09 ~(1.30) c) 3.413 (E +6) BTU /hr X 2652 Mw(t) = 9.0513 (E +9) BTU /hr 9.0513 (E +10) BTU /hr = (flowrate)(1.39 BTU /lbm-F)(67dT) 9.0513 (E +10)/(1.39)(67) = 9.719 (E +7) lbm/hr
[ Answer allowance - between 9.6 (E +7) to 9.8 (E +7)] .
(0.4 - calculation 0.1 - answer) (Errors Carried Fwd) i b) 9.0513 (E +9) BTU /hr = 9.719 (E +7) lbm/hr (Delta h) 9.0513 (E +9) BTU /hr/9.719 (E +7) lbm/hr = 93.13 BTU /lbe
[ Answer allowance - between 90.O' BTU /lba to 95.0 BTU /lbm]
(0.4-- calculation; O.1 - answer) (Errors Carried Fwd) l c) INCREASE (0.5) {
REFERENCE J Fcrley LP OPS-30905C, Obj 4 AND OPS-30905D, Obj 2 and 3 AND
-Qun1 Req'mt Element #CRO-648 KAIR 3.1/3.4 193OO7K108 ...(KA*S) l l
- l'. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE C5 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- FARLEY -98/11/14-REGION II ANSWER 1.10 (2.00)
'Stoca system enthalpy each loop =
- 1199.7 BTU /lbm (0.1)
Fccdwater enthalpy each loop = (-)*405.2 BTU /lbm (0.1)
Enthalpy Rise / Change = '794.5 BTU /lba (0.25)
Loop A
- Power Out = 3.772(E +6) X 794.5 = 2.997(E +9).(0.25)
Loop 3
- Power Out = 3.791(E +6) X 794.5 = 3.012(E +9) (0.25)
Loop C
- Power Out = 3.784(E +4) X 794.5 = 3.OO6(E +9) (0.25)
Total Power Output =
- 9.015(E +9) (0.25) i 9.015-(E +9) BTU /hr
-- =
- 2641.4 MW/2652 MW =
- 99.6 % RTP 3.413 (E +6) BTU /hr/MW (0.5 pts for " BTU /hr-to-MW" conversion - 0.05 pt for answer)
[ Answer allowance - between 99.0 % RTP to 99.9 % RTP]
[Give Full Credit If " Average" FW Loop Calcu Performed]
[Any Errors Carried Forward]
REFERENCE Fcrioy LP OPS-30911; Objectives 1, 2, 4, 5 and 6 AND STP-109.0 AND Steam Tables AND Gual Req'mt Elements #CRO-152 end 648 KAIR 3.1/3.4 3.1/3.4 193OO7K106 193OO7K108 ...(KA*S) m _
s1. t PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 86 THERMODYNAMICS. E AT TRANSFER AND FLUID PLOW ANSWERS -- FARLEY -88/11/14-REGION II ANSWER 1.11 (1.50)
I c) FALSE (0.5) b) FALSE (0.5) i c) FALSE (0,5) ;
REFERENCE Fcr1cy LP OPS-30907B, Objective 1 AND Farley LP DPS-30909, Objcctives 3 and 9 AND Farley LP DPS-30911, Objective 9 AND LP DPS-52101C, Obj ective 5 KAIR 4.1/4.2 OO2OOOK111 ...(KA'S)
ANSWER 1.12 (1.50) o) CLOSED (0.5) b) PUMP SPEED (0.5) c) SERIES (0.5) i REFERENCE Fcr1cy LP DPS-30913 Objective 24 AND Farley LP OPS-30910, Objcctives 3, 6 and 7 KAIR 3.3/3.4 2.2/2.3 3.1/3.3 193OO6K110 193OO6K113 193OO6K115 ...(KA*S) i l
1 1
J
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE G7- ]
TERM 0 DYNAMICS. EAT TRANSFER AND FLUID FLOW !
ANSWERS -- FARLEY -88/11/14-REGIDN II ANSWER 1.13- (1.00)
- USE ATTACHED FIGURE IN GRADING THE QUESTION -
I LOOK FOR THE FOLLOWING: 1 a) Correct placement / labeling of Hi Head Pmp curve. (0.25) b) Correct placement / labeling of Lo Head Pap curve. (0.25) c) Correct placement / labeling of Init Sys curve. (0.25)
'd) Correct placement / labeling of curve depicting f system pressure reduction due to LOCA. (0.25)
REFERENCE Fcrloy LP OPS-30910, Objectives 2, 4, 7 and 123 LP OPS-40302C, Obj. 103 LP DPS-52532H AND Gual Req'at EDP's
-KAIR 2.8/3.1 3.5/3.9
-OO6020K601 OO6000K506 ...(KA'S)
ANSWER 1.14 (1.50) c) -TRWE- FALS 6 ( O. 5 )
b) TRUE (0.5) c) TRUE (0.5)
REFERENCE For1cy LP DPS-30912, Obj ectives 4, 5, 6 & 73 Farley LP :
OPS-30913, Objectives 10 and 11; Farley Technical l Sp;cification 3/4.4.2 LCO and Bases; Farley Technical I Sp3cification 3/4.10.1 LCD and Bases AND Farley Technical Sp:cification 3/4.10.2 LCO and Bases KAIR 2.8/3.2 3.3/3.7 193010K101 193010K104 ...(KA*S)
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE C3 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW i
' ANSWERS -- FARLEY -88/11/14-REGION II i q
i
-ANSWER 1.15 (1.00) o). (1.0)
. REFERENCE
.Fcrloy LP OPS-52105C, Obj ective 2 3 FNP-1-SOP-28.1, " Turbine G:ncrator Operation"; FNP-0-SOP-38.0 3 " Diesel Generators" AND Gual Req'mt Element #CRO-503
.KAIR 2.8/2.9 062OOOA403 ...(KA*S)
ANSWER 1.16 .(1.00) o) (1.0) i REFERENCE Forloy LP OPS-52105C, Obj ective 2 AND FNP-0-SOP-38.0, "Diocol Generators" AND FNP-1--SOP-28.1, "TG Operation" AND Guc1 Req'mt Element'#CRO-503 KAIR 2.8/3.2 ,
062OOOA215 ...(KA'S)
I fi. PRINCIPLES OF NLM FAR POWER PLANT OPERATION. PAGE 99' Tle.r@iODYNAMICS. lEAT TRANSFER AND FLUID FLOW ANSWERS.-- FARLEY -88/11/14-REGION II l
ANSWER 1.17 .(2.50) ,
Ecch.tsorth 0.5 points a) HIGHER (Less resistance to flow >> Other RCPs speed up) b) LOWER (Higher Steam Flow >> P steam decreases) c) HIGHER (Less total flow across core >> delta T increase T-HOT increase witte rods in manual) i d) LOWER (As above, de.lta.T increases, T-COLD decreases) o) SAME (Primary Power = secondary load)
- REFERENCE Forley LP OPS-30913, Objective 15 AND Farley LP OPS-30005C Objcctive-4 KAIR 3.4/3.7 002000A105 ...(KA*S)
ANSWER 1.18 .(1.00) a) DECREASE (0.25) b) REMAIN THE SAME (0.25) j c) DECREASE (0.25) d) INCREASE (0.25) ;
REFERENCE Fcrley LP OPS-30900, Objective 1 AND Farley LF OPS-30905E, Objcetive 4 KAIR -2.6/3.0 010000K502 ...(KA*S) i 1
i' l
L
m 1
4
'1. PRINCIPLES-OF NUCLEAR POWER PLANT OPERATION. PAGE 90 )
I THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS 'FARLEY= -88/11/14-REGION II i
ANSWER 1.19 (2.00) ,
[EACH of the following FOUR (4) worth 0.5 points]
- 1) Rod alignment within a group of control rods is maintained.
- 2) Rods.aea maintained above their insertion limits.
- 3) Rod banks and' overlap ar's maintained per rod insertion. licit technical specifications. l
- 4) AFD is maintained within limits.
NOTE:
If the candidate lists GPTR as a limitation then allow one-half credit (0.25). Total question pt value NOT to exceed (2.0).
1 REFERENCE Forloy LP OPS-52702, Objective 2 AND Qual Req'mt Element CCRO-643 KAIR 2.9 3.3 3.9 193009K107 00iOOOK507 00100GK510 ...(KA*S) 1 i
i f
- 1. 1 L PRINCIPLES OF p_n FAR POWER PLANT OPERATION. PAGE 91 I 1.
THERMODYNAMICS, IEAT TRANSFER AND FLUID FLOW ANSWERS -- FARLEY- -BG/11/14-REGION II I
1 ANSWER 1.20 (2.00) l i
o)' DECREASE (0.5) l b) INCREASE- (0.5) l l
l c) INCREASE (0.5) d)' DECREASE (0.5)
REFERENCE NUS,-Reactor-Operation, Unit'SEVEN (7), Reactor Control, Pcro. 7.4, Pages 7.4-1 through 7.4-5 KAIR 2.9/3.4 3.5/3.7
'OO1000K502 OO1000K509 ...(KA*S)
ANSWER 1.21 (0.50)
TRUE (0.5)
REFERENCE NUS, Plant Performance, Para 1.6, Pages.l.6-1 through 1.6-6, Pero 7.3, Pgs 7.3-1 through 7.3-4 KAIR 3.4/3.9 2.4/2.5 OO2OOOK500 193OO3K102 ...(KA*S) 1.
'2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEriS PAGE 92
- ANSWERS -- FARLEY -88/11/14-REGION II O
]
i ANSWER 2.01 (1.00)
Any EIGHT'(8).of the following - worth 0.125 pts each..
c) Spent Fuel Pool Heat Exchangers g) RCP Oil Coolers b) Charging Pumps Seals h) RCDT Heat Exchanger c) Charging Pump Coolers i) Excess Letdown HX d) RHR Heat Exchangers j) Seal Water HX f o) RHR Pump Coolers k) Letdown HX f) RCP Thermal Barrier Heat Exchanger
.) '
REFERENCE Fcrioy LP DPS-40204A, Objectives 2 and 3 AND Qual Req'mt Elcocnts CRO-631 and 632 KAIR 3.3/3.4 OOZOOOK102 ...(KA*S)
ANSWER :2.02 (1.00)
The caparation assures that both trains of CCW are not occrificed by one supply or surge line failing. . (1.0)
REFERENCE Furany LP OPS-40204A, Objective 6 I KAIR '3.3/3.4 00GOOOGOO7 ...(KA*S)
'2 . -PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 93 )1 1
JANSWERS -- FARLEY -BB/11/14-REGION II i
ANSWER 2.03 (2.00) a)'SAME (0.2) b)-GREATER THAN (0.2) c ) -6CCC TlL^" 'S om E (O.2) d) bESG-TNAN 5^m* (0.2) o) 4 ESS-THAN- 5 6 d (O.2) f)-LCCS THAN SA "C (0.2) i
\
g) SAME (0.2) h) LESS THAN.TNote: "Non e >:isten t" ) (0.2)
- 1) LESS THAN (Note: " Nonexistent") (0.2)
J)-LCCC TliAN f6*SI (O.2)
REFERENCE Fer1cy LP OPS-40101B, Objectives 5 and 11 AND Quel Req'mt
'Elem:nts #CRO-357-KAIR 2.9/3.4 3.7/3.7 ,
076000K403. 076000A302, ...(KA'S) l i
4
1 PLANT' DES'IGN INCLUDING' SAFETY AND EMERGENCY' SYSTEMS PAGE 94 ANSWERS -- FARLEY ~B8/11/14-REGION II >
.)
ANSWER 2.04 (1.00) i
-l Annures a' minimum flow for removing RHR pump heat. (1.0)
NOTE: References to the 500 GPM outo open and the 1000 GPM i auto closure featuresfARE NOT required for full _
l credit, HOWEVER,-if Unit TWO (2) flow rates of 1200 1 '
GPM_and 2000 GPM, respectively, are expressed, this will result in a 0.2 point deduction.
REFERENCE' :
- Fcr1cy. LP OPS-40301K, Objectives 1 and 4 }
KAIR 2.7/3.0 005000K406 ...(KA*S)
, I ANSWER 2.05 (3.00) f.! -'o)'1) Remove iodine'f' rom the containment atmosphere. (0.5) 2)-M6nimize post-LOCA-(general) cor/osion of components by maintaining spray fluid SH of at least B.5. (0.5) l l
b) Pariodic testing of the CS Puups is accomplished. by using i the-(2-inch) recirculation lines back to the RWST. (1.0) c)'1) RWST (0.5)
- 2) Containment Recirculation Sump (0.5)
REFERENCE Fer1Gy LP OPS-40302D, Objectives 4, 7, and 9 AND Qual Req'mt Elements #CRO-654 acid 657 KAYA '4.2/4.3 3.1/3.6 3.5/3.7
- 026000K401 026000K402 026000G007 ...(KA*S) 4 m
_-___m_ - _ . _ _ . . _ _ . _
+
L'
- 2. .. . PLANT DESIGN _INCLUDINE..SAEETY. . AHD_.EMERSENCY.._ SYSTEMS PAGE 95' ANSWERS'-- FARLEY -88/11/14-REGION'II o
ANSWER 2.06 -(2.00) c) Loop A Intermediate Leg (0.25) b) Loop B ' Hot Leg (0.25)
- c) Loop C - Intermediate Leg (0.25)
)
d) .l.oop B1- Cold Leg. (0.25) ]
~l
- o) Loop A - Cold Leg (0.25) l
'f). Loop A - Cold Leg (0.25)
Loop B - Cold Leg ( 0.'25 ) i g)-Loop B - Intermediate Lag (0.25)
\
. REFERENCE l Farloy LP, DPS-40301A, ' Objectives 6, 7, and B 1 KAIR 3.7/4.0 3.5/3.7 4.1/4.1 l OO2OOOK106 .OO2OOOK107, OO2OOOK109 ...(KA*S).
l ANSWER 2.07 (1.50)'
c) 100. percent (0.5) b) 1) Instrument Air System (0.5)
- 2) Main ~ Steam Atmospheric Relief Valve Emergency Air Compressor (0.8)
REFERENCE Forley LP DPS-40201D, Objectives 1, 4 and 8 AND Qual Req'mt Elcaent #CRO-335 .
KAIR 2.6/2.7 3.2/3.6 3.6/3.7 061000K602 061000K502 061000 GOO 7 ...(KA'5) i l
l i
\
N 2. PLANT' DESIGN ' INCLUDING SAFETY AND EistsENCY SYSTEMS PAGE 96 ANSWERS ~-- FARLEY= -88/11/14-REGION II l
1
-ANSWER 2.00- (1.00) c) Provides sufficient NITROGEN volume for complete accumulator discharge. (0.5)
-b) PREVENTS "early" discharge (injection).of water into the core and $njection f ( 0. S ) -
GMart %t %e amonk oOcondensible nitrogen.orz %cdwiGbc)@ed Mb etfECS di4 be mi NOTE: If the following Technica1'Speelfication is given, one-half credit (1.0) should be given:
"The limits en accumulator volume, (boron concentration) and pressure ensure that the i 1
assumptions used for accumulator injection in the safety analysis are met." i REFERENCE Faricy LP OPS-521029, Objective 3 AND Farley Technical ;
Specification LCO and Bases 3.5.1 AND Qual Req'mt Elements CCRO-82, 83 and 84 l I
KAIR 3.4/3.9 006000K602 ...(KA*S) I l
[
ANSWER 2.09 (1.00) a) (1.0) j REFERENCE Ferloy LP OPS-40202A AND Farley LP DPS-52201I, Objectives 7, 12, cnd 14 KAIR 3.6/3.7 045010K111 ...(KA*S) !
'l 1
.. l l
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 97
'l ;
. ANSWERS -- FARLEY -88/11/14-REGION ~II l
l I
ANSWER 2.10 (1.00) < .
]
l'
-Tha valve design limits the plant cooldown rate.... (0.5)
IF cny1DNE (1). valve inadvertently eticks open. (0.5)
I
' REFERENCE Farloy LP DPS-40201A, Objective 7B KAIR 3.1/3.5 035010K602 ...(KA*S)
ANSWER 2.11~ (1.00) I c)- TWO (2) (0.5) b) To reduce the site boundary dose that would occur if a single tank ruptured. (0.5)
. REFERENCE Fcr1cy LP DPS-52106B, Objective 3 AND FNP-1-SOP-51.0, " Waste Geo System" procedure AND Technical Specification LCO and Decas 3/4.11.2.6 AND LP OPS-40303B, Objective 3 KAIR 2.5/2.8 2.5/2.7 071000 GOO 7 0710006010 ...(KA*S)
. ANSWER 2.12 (1.00) c) (1.0)
REFERENCE Farley LP DPS-52101K, Objective 11 KAIR 3.4 3.4 2.9 ,
OO5000A402 OO5000K402 OO5000K403 ...(KA'S) i i
i
'1 PAGE 93 2 .~ PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS ,
ANSWERS -- FARLEY' -BS/11/14-REGION II
\
1 4
- - 1 i
ANSWER 2.13 (1.50) a)' Acts as a " backup" to the hydrpgen repombiner subsystem. ]
, , by L%cm4M3 t4 2 frot, We cordaimar (A4l AdmW( Ap (0.5)
P.c j kyk o$en -f< ee co r, <
b)L Ensures that (post-accident) hydrogen concentrations are uniform throughout containment. (0.5) c) TRUE (0.5) l
-REFERENCE Fer1cy LP OPS-40302E,' Objectives 3 and 6 AND LP OPS-40302A KAIR 3.3/3.5 .3.2/3.3
'028000 GOO 4 028000 GOO 7 ...(KA*S)
ANSWER 2.14 (1.00) b) (1.0)
REFERENCE Fcr1cy LP OPS-40204F, Objectives 1, 2, and 4 AND ,
J FNP-1-GOP-36.4, "120 A.C. Distribution Systems" KAIR 2.4/2.9 3.0/3.2 1 062OOOK409 062OOOGOO7 ...(KA'S) l l
1
I. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 99 ANSWERS -- FARLEY -BS/11/14-REGION'II
- j. SD ANSWER 2.15 4 2.00)-
1
- c) . Water Treatment Plant (deoxygenated makeup) Water (0.5) b) Hotwell Condensate (via the condensate pumps) (0.5) c) Demineralized Water Storage Tank (0.5)
, - .d) S vice-Water -@,4F-
[ do g.\ fo,ke cR Or dowr% GO*r'&
REFERENCE ,
l LFerlcy LP OPS-40201D, Dhjective 11 AND FNP-1-SGP-21.0 l I
L" Condensate and Feedwater' System" AND Technical
'Sp cification LCO and Bases 3/4.7.1.3 AND Qual Req'mt SOP's KAIR '3.9/4.2 !
i 061000K401- ...(KA*S) l L. .
, . .. . i.
'2 . PLANT DESIGN INCLUDING SAFETY AN M PAGE X100 ANSWERS -- FARLEY -88/11/14-REGION II j
'l 1
J l
1 ANSWER 2.16 (2.00)
- 1) Only ONE (1) RCP is to be started at any one. time. (0.5) l
.2) After any. running period OR ... (0.1) ,
after any attempted start that fails... (0.1) l allow a minimum 30-minute idle period before attempting a restart. (0.3)
- 3) DO NOT exceed THREE (3) starts or attempted in any f TWO (2) hour period. (0.5) l 1
- 4) IF THREE (3) starts or attempted starts have been i acde within a TWO (2) hour period.... (0.2) allow a 60 minute idle period before an additional otart. (0.3)
REFERENCE Quel Req'mt SOP-1.1 Precautions and Limitations AND FNP-1-SOP-1.1 " Reactor Coolant System' (Limits / Precautions)
KAIR 2.6/2.9 3.3/3.6
.OO3OOOK614 OO3OOOG010 ...(KA*S) l l
l 1
- h. PLANI _ DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE X101 ANSWERS -- FAPLEY -88/11/14-REGION II l
l l
ANSWER 2.17 (2.00) 1 o) (NOTE: 0.17!6 pts for each correct answer)
FCV-113A FCV-113B FCV-114A FCV-114B k '
" DILUTE" (CLOSED) (CLOSED) (OPEN) (OPEN)
" BORATE" (OPEN) (OPEN) (CLOSED) (CLOSED)
" ALTERNATE" ]
DILUTE' (CLOSED) (OPEN) (OPEN) (OPEN) l b) Volume Control Tank (VCT) Level (0.5) :
REFERENCE '
Ferloy LP OPS-40301G, Objective 4 AND LP OPS-52101G, Objcctive 5 and Figure TWO (2)
KAIR 3.1/3.1 3.0/3.2 OO4000K106 OO4010A105 ...(KA*S) l l
_- -----_-____ _ _ - _ a
L,.
~
- 2. PLANL DESIGN INCLUD,tNG SAFETY AND EMERGENCY SYSTEMR PAGE X102 1 ANSWERS ~-- FAhLEY -88/11/14-REGIDN II l
t
- ANSWER 2.18 (3.00)
[Any EIGHT (8) of the following TEN (10) . worth 0.25 points each.]-
c)' 1) Containment Isolation System i) Containment Spray System d
- 3) Containment Fan Cooler System
- 4) Containment Air Purification and Cleanup System
- 5) _ Habitability System
- 7) Residual Heat Remcval System
- 0) Combustible Gas Control System
- 9) Penetration Room Filtration System i
- 10) Auxiliary Feedwater System h) 1) Component Cooling Water System (0.25)
- 2) Servics Water Systein (0.25)
- 3) River Water Oystem (0.25)
- 4) Emergwncy Power System (0.25)
REFERCNCE Ferloy LP OPS-40302A, Objective 2 AND Farley LP OPS-52102J, Objcctive 2 KAIR 3.7/3.0 3.9/3.8 013OOOGOO4 013000G007 ...(KA'S)
I 1
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l l
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-i
- 2. PLANT DESIGN INCLUDING SAFETY'AND EMERGENCY SYSTEMS PAGE X103 ANSWERS -- FARLEY -88/11/14-REGION II ANSWER 2.19 (2.00) c)- Blowdown - starts with initiation of the LOCA and ends when'the reactor coolant system pressure falls to that i of the containment. (0.5) 1 b) Refill - commences at the end of the blowdown and ends when the addition of emergency cooling water fills the bottom of the reactor vessel and reaches the bottom of the fuel rods. (0.5) c) Reflood - the time from the end of the refill phase until the reactor vessel has been filled with water to the extent that the core temperature rise has been terminated and the core temperatures have been reduced to the long-term values associated with decay heat removal. (0.5)
?
d) Long-Term Recirculation - the time decay heat removal cystemslare " switched" to the containment sump (s) and cource of cooling medium is recirculated within
-containment. Temperatures are maintained at the long-term values.
NOTE: Reasonable Wording Accepted REFERENCE Forlay LP OPS-52702, Objective 5 AND 10 CFR 50 App. K
.information KAIR 4.0/4.1 3.8/4.2 000011K305 000011K313 ...(KA*S)
-1
. t >
.3.- INSTRUMENTS AND CONTROLS PAGE X104 ANSWERS -- FARLEY -88/11/14-REGION II
'i 2 WOO ANSWER 3.01 40. 0^ F i c) Low Pressurizer Pressure Safety Injection (SI) (0.5) b) 1) Main Steam Line Isolation (0.5)
-2) Lew "eir. St::a Lin Fressure 51-- (0.5)
(cl0 noi ded4ct f ol d s i f 'L oD o ^ p is e m e n W )
c) The7 rain A and Train B steam dump interlock switches are placed in the " BYPASS INTERLOCK" position. (1.0)
REFERENCE For1cy LP OPS-522011, Objective 14 and 22; LP OPS-40301E; LJS OPS-52201H, Objective 2 AND LP-52201G, Objectives 6 and 13 KAIR 3.0/3.3 3.7/3.9 041020K409 013OOOK412 ...(KA*S) l l
)
I 1
I i
l l
1
.h. INSTRUMENTS AND CONTROLS PAGE X105 ANSWERS -- FARLEY. -88/11/14-REGION II l l
' ANSWER 3.02 -(1.50) i a) Cany FOUR'(4) of the following FIVE (5) " signals" i worth (0.25) points each] .l
. .. 1
- 1) Low 011 Pressure ]
- 2) HP Air Temperature
- 3) High 011 Temperature-
- 4) Cooling Water Drain High Temperature 1
- 5) High Bleed-off Pressure Switch b) FOUR (4) starts (0.25) - 15 minutes (0.25) 1 J
REFERENCE l Ferloy Qual Req'mt Elements CRO-572, CRO-577 and CRO-579, !
.For1Gy'LP OPS-52100A, Objective 2 LP DPS-40204D, Obj ective 4 AND FNP-1-SOP-31.0 " Compressed Air System" l KAIR 2.4/2.6 2.9/3.0 2.8/2.9 i 078000K601 078000 GOO 7 078000G010 ...(KA*S)
ANSWER 3.03 (1.00) !
d) (1.0) {
i REFERENCE Farloy Qual Req'at Element CRO-147, Farley LP OPS-52201D, Objcctives B, 21 and 24 KAIR 3.7/3.8 2.7/2.9 l 015000K407 015000K502 ...(KA*S)
I i
i t
i
_ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ . _ . _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ . _ . _ _ _ _ I
4 .
~3. INSTRUMENTS AND CONTROLS PAGE XiO6 i ANSWERS -- FARLEY- -80/ii/14-REGION II l
l 1
l ANSWER 3.04 (3.00) H
[ Permissive worth 0.33, Logic - 0.33, Function - 0.33]
c) P-G (Single Loop Loss-of-Flow Permissive) - 2/4 -
Prevents Rx Trip from loss of flow (or RCP Bkr open in single loop) below a preset value of 35 percent power. ,
b) P-9 (Turbine Trip Permissive) - 2/4 - Prevents a Rx Trip from a turbine trip when below a preset value of 35 percent power.
c) P-10 (P/R Power Escalation Permissive) - 2/4 - Allows I
for the blocking of the I/R Hi Flux Rx Trip, the I/R Hi Flux Rod Trip and the P/R Hi Flux (Low Setpoint) Rx Trip AND interlocks w/ S/R Hi Volt - inhibits manual reset.
NOTE: P-7 is an INDIRECT permissive, however, if listed, allow for 1/2 credit - i.e.; 0.160 for the permissive, the logic and the description.
REFERENCE Fcrisy LP OPS-5220iD, Objective 18 KAIR 3.7/3.8 015000K407 ...(KA*S)
7_
.. t
- 3. INSTR'JMENTS AND CONTROLS PAGE X107 ANSWERS --'FARLEY -88/11/14-REGION II ANSWER 3.05 (2.00)
)
c) [Any FOUR (4) of the following FIVE (5) conditions - )
each worth O.25]
- 1) Regulation Failure l i
- 2) Phase Failure l 3)~ Logic Error I
- 4) Multiplex Error
- 5) Loose or Removad Printed Circuit Card 1 Vr.a c_ (4) !
b) [Any TWO (2) of the following THREE-(3), actions - l each worth 0.25] i
- 1) Stops ALL rod motion when in AUTO or MANUAL l
- 2) Orders due:d current to associated stationary and ,
movable gripper coils i I
- 3) Overrides any orders from Logic Cabinet for the !
associated Power Cabinet V) in u i n iv s hse t c) TRUE REFERENCE Fcr1cy LP DPS-52201E, Objectives 12, 13, 14 and 15 KAIR 3.7/3.9 OO1050A201- ...(KA'S)
1
,. ~' 3 . INSTRUMENTS AND CONTROLS PAGE-X108 I!. . f _.
' ANSWERS -- FARLEY -88/11/14-REGION II'
[ ANSWER 3.06 (2.00) a) (Either of the answers below worth 1.0 pts - NOTE:
Reasonable wording is acceptable.]
The level program signal originates from P-imp. The l turbine pressure supplies the ZERO (0) percent to 100 parcent power reference.
OR . . . ,
The turbine impulse chamber pressure signal develops the desired S/G 1evel program for all powar' levels.
b) 1) INCREASE (0.5)
- 2) The failed high steam pressure trknamitter causes the steam flow input to SGWLC to intrease. (0.5)
REFERENCE Farley LP DPS-52201B, Objectives 1 and 2 (Second Sentence)
KAIR 3.6/3.8 3.4/3.6 035010K401 035010A203 ...(KA'S) ;
l l
G. INSTRUMENTS.AND.. CONTROLS -PAGE X109 ANSWERS -- FARLEY -88/11/14-REGION II q i
~l i
ANSWER 3.07 (1.50)
[Each parameter worth 0.5 pts) c) Total Steam Flow (Summation of FT-474, 484, 494 for channel III signal or summation of (
F.T-475, 485, 495 for ch IV signal) l b) Steam Header Pressure (PT-464) 1:) Feedwater Pump Discharge Header Pressure (PT-500)
NOTE: Listing of channels IS NOT required for. full credit, however,.if incorrect listing of transmitter / channel is made it should result in a point deduction.
REFERENCE Fer1cy LP. OPS-52201B, Objectives 1 and 4 KAIR 3.4/3.4 3.6/3.8 059000K104 035010K401 ...(KA*S) l l
l I
A-
'3 . INSTRUMENTS AND CONTRDLS PAGE'Y.110' 1
ANSWERS -- FART.EY -88/11/14-REGION II !
l l
ANSWER 3.08 , (2.00) i a) Charging reduced upon receipt of' level error... (0.25) l Level initially decreases... (0.25) ,
Level reaches 15 percent, L/D isolates and !
Minimum charging (seal inj) continuez..... (0.25)
Level increases until reactor trip. (0.25) !
in crea 64 do thf -
.b) PZR level errin: the seme...
(0.25) .
Maximum reference level setpoint i; the... (O IS) 'I
-- r :s the 1:v;l ::tpcint et 100 p;r; nt RT.". (0.25) :
c) PZR is unaffected by the' failure... (0.25) l PZR level program is determined by Actioneered i HIGH T-AVE and-" set" values of level - i.e.g Turbine 1 Impulse Pressure has nothing to do with it. (0.25) {
NOTE: Reasonable wording acceptable REFERENCE Ferloy LP- DFS-52201H, Objectives 6 and 9 KAIR 3.2/3.4 3.4/3.6 3.1/3.3 011000K301 011000A210 011000A104 ...(KA*S) !
13'. -INSTRUMENTS AND CONTROLS PAGE X111 s
ANSWERS -- FARLEY -88/11/14-REGION II i 1
.ANSWERJ 3.09 (2.50) -
o) The " mode selector switch" must be in the MODE ONE (1) or MODE TWO (2) positions, AND ... (0.5)
-The " mode FOUR (4) selector switch" must be in the OFF position. (0.5) b) [Any 6.of the following 7 - each worth 0.25 pts]
- 1) Local Stop Push Button (Manual Stop - Local) ,
i
- 2) Remote Stop Push Button'(Manual Stop - Remote) l
- 3) Lube Oil High Temperature
- 4) Jacket Water High Temperature I
- 5) Jacket Water Low Pressure
- 6) Crankcase High Pressure
- 7) Barring Device Engaged (" Big" Diesels) {
REFERENCE Ferlay LP OPS-40102C, 0bjectives B, 10 and 11 AND For10y LP DPS-52102I, Objective 1 KAIR 4.1/4.4 3.8/4.1 3.9/4.2 013OOOK112 064000K401 064000K402 ...(KA'S)
I l
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- 5; INSTRUMENTS AND CONTROLS PAGE X112-ANSWERS -- FARLEY -88/11/14-REBION II i
ANSWER 3.10 (1.00) w iT-I-O^-UN:T :: pel t.lu,, _ (0.25; Only-UV-eignals end 4 L M n=1s~ f en= - the-4elec4edei-L wi l l utart the diesel- 'n 5)
UNIT 'I AND UNIT II position LA4HS7 -- Signals from either b3D) unit will start the diesel. . jar &T REFERENCE Faricy LP OPS-40102C, Objective 8 Fer1cy LP OPS-52102I, Cbjective 1 KAIR 4.1/4.4 013000K112 ...(KA*S)
Ullibf d ri4 ton - 9/Li aka3 .- (4,9- / 5,pc .ty, c.33; YnNZ PoS;% - Djn ,wa,.s. 0A 2 W1 '"/ 6 53) j
1 L
- 3. INSTRUMENTS AND CONTROLS PAGE X113..
ANSWERS -- FARLEY -89/11/14-REGION II
' ANSWER 3.11 (2.50).
lc) Bank DNE (1) . starts.toJopen at *FOUR (4) degrees delta T rods - WHEREAS.
forthe" Loss-of-Load" controller-thisallowsfor)..
initial load rejection response fro
' Bank DNE (1) immediately opens *ZERO (0) degrees delta T for the " Turbine Trip controller. (1.0) b) Cath the " Loss-of-Load" and " Turbine Trip" controllers have a TEN (10) delta T error (mismatch) setpoint. (0.5) c) The " Loss-of-Eoad" controller has a 16 degree F delta T crror (mismatch) setpoint WHEREAS...
the " Turbine Trip" controller uses a 20 degree error cetpoint. (1.0)
REFERENCE I Forloy LP- DPS-52201G, Objectives 4, 6, and 9 KAIR 2.3/2.6 2.8/3.1 2.5/2.8 .3.4/3.6 041020K403 041020K411 041020K414 041020K418 ...(KA'S)
I i
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~ '3 . INSTRUMENTS AND CONTROLS PAGE X114 I
- m. l ANSWERS'-- FARLEY ' -88/11/14-REGION II 1
1 ANSWER 3.12 (3.00) betge\t #
h Interlock worth 0.33; Logic /Condi - 0.33; Des tion - 0.33).
1 c) C-7 (Loss o Lead Interlock)'- I 1se Pressure signal 1 from PT-447 (NOTE (Nostep "real" I ic associated with this interlock) 10 percen o 5 percent / min ramp.- Arms oteam dumps in T-AVE (mMp _
f ter load rejection, IF C-9 is present.
b) C-8 (Turbine Trip I erlock) -
bine Trip indications; i.e. 4/4 Stop Va es closed OR 2/3 to Stop 011 signals
- Arms steam d ps in T-AVE mode afte turbine trip, IF C-9 is prese .
c) C-9 (C enser' Interlock) - Circ Pump BKR closed and >B inch Hg condenser vacuum - Prevents overpress ization en subsequent damage to the main condensers.
REFERENCE Fcrioy LP OPS-52201G, Objectives 6, .and 9 KAIR 2.5/2.8 2.4/2.7 2.8/3.0 041020K414 041020K407 0410200007 ...(KA*S)
- 1) a) C-7 (Loss of load interlock) (.33) b) C-8 (Turbine trip interlock) (.33) l c)_ C-9 (Condenserinterlock) (.33)
I Locic/ Condition
- 2) a) 15%/120 seconds (.33) b) 4/4 stop valves closed or (.33) 2/3 low auto stop oil pressures c) Circulating water pump breakers closed and condenser vacuum < 8 in. Hg vacuum (.33) ;
- 3) a) Arms steam dumps in Tavg mode after (.33) load rejection b) Arms steam dumps in Tavg mode after turbine trip (.33) l c) Prevents steam dump valve operation when they could cause ovegressurization and damage to the main condenser (.33)
l
- 3. t INSTRUMENTS AND CONTROLS PAGE X115 i
ANSWERS -- FARLEY -B8/11/14-REGION II
~ ANSWER 3.13 (1.50) .
c) . Trolley Controls (0.5) ;
b) Eridge Controls (0.5) c) Holst Controls (0.5)
REFERENCE l Fcricy LP OPS-52iOED, Objective 1 AND Qual Req'mt Element CCRO-234 KAIR 2.5/3.0 3.0/3.0
.034000 GOO 7 .-034000 GOO 9 ...(KA*S) i
)
ANSWER 3.14 (1.50)
[Any THREE (3) of ths following FIVE (5) - worth O.5 g points each3 {
r I
c) The "ON-OFF" selector switch in containment must be j selected to "ON". ]
b) The transfer tube gate valve must be opened.
c) The upender frame - reactor side - must be in the DOWN j position. i i
d) The upender frame - spent fuel pool side - must be in ]
the DOWN position.
o) The s:Irt push button must be momentarily depressed.
REFERENCE Fcrisy LP OPS-52108D, Objective 9 AND Gual Req'mt Element CCRO-235 KAIR 2.5/3.3 2.5/3.0 3.0/3.0 034000K402 034000 GOO 7 034000 GOO 9 ...(KA*S)
3. INSTRUP9ENTS AND CON 1RfM.S PAGE X146
- . ANSWERS -- FARLEY -BG/11/14-REGION I'I L.s
'I,.
' ANSWER 3.15- ( 1. 50)'
c) 'The' Fuel Bldg supply and exhaust fans trip. (0.5) b) .The Fuel Bldg HVAC supply / exhaust dampers close. (0.5)
- c) 'All FOURl(4) Penetration Room Filtration Exhaust And
' Recirculation Fans auto start. (0.5)
REFERENCE ~
l 1 Fcrley LP DPS-40305A, Objectives 5 and 6 AND Qual Req'ot l Elem3nt " Misc. #13F"
' KAIR ' 3.2/3.4 2.9/3.1
. 072OOOK402 072OOOA301 ...(KA*S)
ANSWER 3.16 (1.00)
- 4) At 2000 PSIG,'if.the valves are.not open, an alarm counds on-the MCB.. Once' acknowledged, the alarm will ,
reflash at on at ONE (1) hour intervalc, if the valve IS NOT capable of being opened., (0.5) )
=l b). If the valve is capable of being opened, it will automatically open at 2000 PSIG. (0.25) c) An "S"-signal automatically opens these valves should -
they be closed for maintsnace when a safety injection is ]
initiated. (0.25) q l
.1 REFERENCE l Forlay LP OPS-40302C, Objectives 7 and 10 AND Qual Req'mt Eloments #CRO-626 and 627_
KAIR 3.2/3.5 4.0/3.9 2.9/3.4 OO6000K412 OO6000A301 OO6000k419 ...(KA*S) l l
1
4
- 4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 7.117 j 4 RADIOLOGICAL CONTROL 1 ANSWERS -- FARLEY -88/11/14-REGION II ;
I ANSWER 4.01 (1.50) o) Lssa than or equal to 25 Rem Whole Body (0.25)
Loss than or equal to 3OO. Rom Thyroid (From Iodine)(0.25) b) Low Population Zone (LPZ) - Individual located on boundary WILL NOT exceed guideline values if EXPOSED TO CLOUD DURING THE ENTIRE PASSAGE OF THE CLOUD. (0.5)
Exclusion Area - Individual located on boundary WILL NOT !
cxceed guideline values for TWO HOURS IMMEDIATELY FOLLOWING POSTULATED FISSION PRODUCT RELEASE. (0.5) i REFERENCE
-Fer1cy LP OPS-40302A, Objective 4; Fe.imy LP OPS-52102J, Objcctive 43 Farley Technical Specification LCO and Bases 3.4.7.2 AND 10 CFR 100 KAIR 2.4/3.9- 3.1/3.9 !
OOOO38GOOO 000060G004 ...(KA'S)
ANSWER 4.02 (2.00)
"A single failure of any active component, or a single failure of any passive component DOES NOT (WILL NOT) result in a loss of the capability of the electrical or fluid system-to perform its intended safety function (essuming that): (1.0)
- 1) Any multiple failure (or failures) resulting from a single occurrence are considered " single failures",
AND ...g (0.5)
- 2) If an active failure has occurred, passive components will function properly and if a passive failure has occurred, active components will function properly.
(0.5)
NOTE: REASONABLE WORDING ACCEPTED l REFERENCE l
Fcr1cy LP OPS-40302A, Objective 33 Farley LP OPS-52102J, Objtetive 3 AND 10 CFR 50, Appendix A KAIR 3.1/3.4 2.9/3.3 3.7/3.8 013OOOK400 013OOOK502 013OOOGOO4 ...(KA'S)
7
']i 4i PROCEDlEEct - NORMAL. ABNORMAL. EMERGENCY AND PAGE X113 j RADIOLOGICAL.. CONTROL ,
l j ANSWERS -- FARLEY -88/11/14-REGION II l i
1 ANSWER 4.03 (1.00) '
.c) (1.0) l '
REFERENCE
~FNP-O-AP-16, " Conduct of Operation - Operations Group"3 10 CFR 50.54(1), 10 CFR 55.4 and 10 CFR 55.13 AND Fi;P
-Op; rations Memorandum 81-02 KAIR 2.5/3.4
-194001A103 ...(KA*S) i l
- a
9
- 4. PROcKDURES - NORMAL. ABNORMAL.'EnnGENCY AND PAGE X119 ftADIOLOGICAL CONTROL ikNSWERS - FARLEY -88/11/14-REGION II ANSWER 4.04' (1.00)
'Ths following assumptions must be expressed / mentioned:
a)' Nn deviation.from the original INTENT of the procedure was taken. '(0.5) b) Approval of the change.was apparently made by TWO (2) !
members of-the plant staff, one of which is a licensed SRO. (0.5)
. REFERENCE FMS Technical Specifications, Sec. SIX (6), Pgs 6-12, 6-13 KAIR 4.1/3.9 194001A102 ...(KA'S)
ANSWER 4.05 (2.00) c). Start Boric Acid Transfer Pump 1A or 1B (0.5) b) Open Emergency Borate to Charging Pump Valve (0.5) or opea eevt Sup{
c) Verify ONE (1) Charging Pump running (0.5) !
d) Verify Charging Pump Suction Header Isolation Valves are open (0.5) or verih open 8lJJ A, 61310 ;
REFERENCE
'Fcrisy LP OPS-52533A; Qual Reqm't Element #CRO-7663 FRP-S.1, "Racponse To Nuclear Power Generation /ATWT" AND i FNP-1-ADP-27.0, " Emergency Boration" l KAIR 4.2/4.4 000024K302 ...(KA*S) i l
l l
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_g j _ - - - _8 . _ . . I
' I a g.a t a 1-a
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a a a s a I
,' 8 a I !
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g l 8 e a ; y I
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= u.m
] '
M r i -
I 8 1
.P ..ar.u Run -
ri i "E .
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- 1 8 m .
-y 98 4 486 I
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ALT n - *
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- c. t : - nllllir
- s. .. , i .
c b
!! O#N m- k..I J L- ,9gsn u-pt
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, 18eSTRUndENT AACK8 -
a
- e ,{ { , Cf3 M
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, .p , . . . . .
gig,1 p ggy,
/
== = " 79CHNICA L SUPPCRT CENTER a a a CONPINES OF CONTROL ROOM
": CONTROLSAAEA ALTdRNATE TECHNICAL SUPPORT CENTER FIGURE 4-1
)
1
- 4. PROCEDURES'- NORMAL. ABNORMAL. Eint-rusENCY AND PAGE X120 RADIOLOGICAL CONTROL ANSWERS - 'FARLEY -88/11/14-REGION II
. ANSWER 4.06 (2.00)
-Sao Key - Each correctly identified area worth 0.20 points REFERENCE Fer1cy LP OPS-53OO2, Obj ective 3 AND FNP-0-EIP-O KAIR 3.1/4.4
.194001A116 ...(KA*S) 1 ANSWER 4.07 (1.00)
The containment LEAKAGE RATES are outside.of the Tech Spec ;
~11aits - (as identified in Technical Specification 3.6.1.2) i (1.0) 4 NOTES: )
Warding in parenthesis NOT REQUIRED for full credit. j Rafcrences made to "other" symptoms is inappropriate and chould be judged " incorrect" since the loss of containment integrity can be directly attributable to no other symptom th!:n " leakage" - the ONLY item truly "affected" is the SOLID SURFACE of the containneent - a " hole" was created by'the ;
eicoile (CRDM). l REFERENCE Fcr1cy Technical Specification Definition 1.6 KAIR 3.8/4.2 3.7/4.3 3.1/4.4 OOOO69A201 OOOO69K301 194001A116 ...(KA*S) l i
1
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i
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)
'+ 4. PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE X121 L
FtADIOLOGICAL CONTROL
-ANSWERS - FARLEY_ -88/11/14-REGION II .,
l i j
ANSWER 4.00 (1.50) o) TRUE (0.5) b) TRUE -(0.5) 3
'c) TRUE (0.5)
]
REFERENCE-Ferloy Qual Req'mts, Abnormal Op Procedures, Farley .
Tcchnical Specification LCOs 3.7.11.3 and 3.7.12, '
FNP-0-AP-39, " Fire Patrols and Watches" AND Operations Stcoding Policy Book, 10-21-06 Memo, "Firewatch Instr."
'KAIR 3.5/4.2 2.9/3.9 194001K116 OOOO67A215 ...(KA*S)
ANSWER 4.09 (2.00)
[Any FOUR (4) of the following FIVE (4) actions - each worth 0.5-points]
'c) Shutdown turbine as expeditiously as possible - IF the Shift Supervisor deems it necessary, TRIP the turbine.
b) Shutdown reactor as expeditiously as possible - IF the Shift Supervisor deems it_necessary, TRIP the reactor. j c) . Verify reactor trip bkrs/ bypass bkrs are open st the e reactor trip breaker cabinet.
d) Verify main turbine tripped at the main turbine governor cnd pedestal.
o) Announce (via PA system) " Fire in the Control Rooms chifting control to the Hot Shutdown Panel." j REFERENCE For1cy Abnormal Op Procedure FNP-1-ADP-28.2, " Fire In The Control Room"; Farley LP DPS-40?O4G, Objective 1 AND " Design Bccoc" Hot Shutdown Panel AND AOP Qual Req'mt ( AOP-28.2) ) '
KAIR 3.3/4.1 4.1/4.5 4.2/4.5 OOOO67K304 OOOO6BK312 OOOO6BK318 ...(KA'S) l l
n-_ _ - _ . ._ _.
' A4'. PROCEDURES - NORf4AL. ABNORMAL. EMERGENCY AftR PAGE X122 RADIOt,0GICAL CONTROL ANSWERS -- FARLEY ,
-88/11/14-RT9 ION II ANSWER 4.10 (2.00)
- c)t Oc... , 1=ve-1-changes--eh reater ui.n c; .:::=1 te -( 0. i'25 F F I'.'E ( 5-)-persen t- RT* te . . . . . (-Ort 25t
' A ,,y mg$4 4ade N 7 ove .rfdwdto^ (. f)
Power 1sv4 che.nges-of-less-tt1armr equrt-te -(0.125T N Y Nr rv---FIVE-t95 F pereen t-RTPv (Ov1251--- L b) Parcant Rated Thermal Power (RTP) (0.5) c) 1). Power Level Reduced FROM (0.25)
- 2) Power Level Reduced TD (0.25)
- 3) Time of the Reduction (0.25)
- 4) Reason for the Reduction (0.25)
' REFERENCE
' Operations Memor.nndum 84-03 (3/3/84) AND Admin Proc. Qual Rsq'mts, Element CRO-399 AND AP-16, Section.6.1 !
KAIR 3.4/3.4 .3.6/3.5 194001A106 192OOOK118 ...(KA*G) t u __--_- _ _ _ - -
- 4 . PROCEDURES - NORMAL. ABNORMAL. EMERGENCY'ANE PAGE '/.123 8 RADIOLOGICAL CONTROL j ANSWERS -- FARLEY -88/11/14-REGION II i
l ANSWER 4.11 (1.50) a) The "B-Schedule lists all STP's which must be performed overy 31 days or more often - i.e.g any surveillance.
frequency greater than or equal to a " Monthly" activity.
(0.5) b) .The "Due Date" will be the EARLIER of ... (0.1) i
- 1) the date last performed PLUS the surveillance interval listed in the " Tech Specr" OR.... (0.2)
- 2) the date previously scheduled PLUS the surveillance interval listed in the " Tech Specs" . (0.2) c). The " Grace Period End Date" is be the EARLIER of..(0.1) !
- 1) the date last performed PLUS the surveillance interval listed in the " Tech Specs" PLUS a " grace ;
period (as allowed by Section 4.0.2 of the " Tech ,
Specs") OR.... (0.2) l 2)'the date performed TWO (2) cycles in the past PLUS THREE (3) surveillance intervals as listed in the
" Tech Specs" PLUS a " grace period (um allowed by Section 4.0.2 of the " Tech Specs") OR.... (0.2)
NOTES: Reasonable Wording Accepted The statement "as allowed by Section 4.0.2 of the " Tech Specs" IS NOT required for full l credit.
REFERENCE j Ferley Technical Specifications Sections 4.0.1 through 4.0.5, FNP-O-AP-5, " Surveillance Program Administrative Control", FNP-O-AP-52, " Equipment Status Control And Maintenance Authorization AND Farley Gual Req'mt, Element CRO-378 KAIR 2.5/3.4 3.6/3.1 194001A103 194001AiOO ...(KA*S)
- - _ - - - _ - _ _ _ - - _ - _ _ _ _ - _ _ _ _ _ __ - i
- 4 PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE X124 l RADIOLOGICAL CONTROL ANSWERS'-- FARLEY -88/11/14-REGION II ANSWER 4.12 (1.50)
CAny SIX (6) of the following SEVEN (7) areas - each worth O.25 points]
c)- RHR Pump Rooms
-l b) Charging Pump Rooms c) Aux Feedwater/ Plant Heating Equipment Room d) Battery /DC Switchgear Room
-o) ' Main Steam Valve Room f) Vital Equipment Switchgear Breaker Panels g) Security. Computer Circuity Panels ;
REFERENCE Farley Administrative Procedure FNP-O-AP-42, " Access
-Control" AND FNP-0-AP-16, " Conduct of Operations - t Operations Group", Sections 4.15 and 4.19 KAIR 3.1/3.4 194001K105 ...(KA*S) i l
l 1
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.-_. m "4 PROCEDURES - NORMAL. ABNORMAL. EPERGENCY AND -PAGE X125 ,
RADIOLOGICAL CONTROL f
..- t
' ANSWERS -- FARLEY -88/11/14-REGION II ANSWER 4.13 (2.00) ,
c) 1) POWER RANGE INSTRUMENTS INDICATE THREE (3) PERCENT INTERMEDIATE RANGE STARTUP RATE SLIGHTLY POSITIVE i
[ Highest Priorityg " Prompt (Orange) Action") (0.5) j l
- 2) NC SYSTEM SUBCOOLING'IS *ZERO (0) DEGREES F -
l FIFTH HOTTEST CORE EXIT TC AVG TEMP '715 DEGREES F
[2nd Priority] (0.5) l NOTE: Priority of' actions NOT necessary for full credit, HOWEVER, IF core cooling is-expressed as a higher ,
priority " prompt action" deduct 0.25 points.
" Heat Sink" is a " Yellow" CSF condition k,
b) A) Proceed with functional restoration of the " loss of subcriticality condition" THEN continue with restoration of the " loss of core cooling" condition.
(1.0)
REFERENCE For1cy LP OPS-52534, Objectives 1, 2, 3 and 4; FNP-1-CSF-0, ;
" Critical Safety Function Status Trees" 1 KAIR 4.4/4.7 4.0/4.4 4.4/4.6 i OOOO29K312 OOOO74K311 OOOO54K304 ...(KA*S) i ANSWER 4.14 (2.00)
- 1. funi cladding (0.5) j
- 2. fuwl matrixhte or h (0.5)
- 3. reactor coolant system (boundary / piping) (0.5) l I
- 4. reactor building / containment (0.5) l l
- NOTE: NO CREDIT to be given for " distance" if listed as one l' of the " barriers" maintained (not CSF related).
REFERENCE I Forloy LP OPS-52702, Objective 1 AND Farley LP OPS-52102J, )
Objtetive 1 and 8 AND LP OPS-40302A, Objective 8 !
KAIR 4.1/4.3 l 002000G015 ...(KA*S)
E_ _ __ _ _
4
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'"4.. PROCEDURES _*'ORMAL. ABNORNAL. F M f E10Y AND PAGE X126 l B M 1 Ql,.Q glI C A L t CONTROL i
ANSWERS -- FARLEY -88/11/14-REGION II l
' ANSWER 4.15 (1.00)
[ANY SIX (5) of the following ~ worth 0.25 pts each)
- 1) Definitions '6) Design Features
- 2) Limiting Safety System '/ ) Admin Controls Settings (Safety Limits)
- 3) Limiting Conditions For 8) Appendix D -Environmental Operation (LCOs) Protection Plan
- 4) Surveillance Requirements 9) Tech Spec Interpretations
- 5) Bases - Sactions'3.0/4.0 10) Bases - Section 2.0 0-the r pard 5" .frt>c T[5i also 4W/ Mfd 45 REFERENCE, Faricy. Technical Specification LCO & Baseu 2.1.1 and 2.1.2 AND Administrative Controls 6.7.1 KAIR 2.6/3.1 2.9/4.0 2.9/4.1 19400J.A108 006000G006 012OOOGOO6 ...(KA*S)
_ _ _ _ _ _ _ _ _ _ _ . - - - - - - 4
. PROCEDURES - NORMAL. ABNORMAL.-EMERGENCY AND PAGE 'X127
,- 88DlQl.QElfdL_ CONTROL ANSWERS -- FARLEY -88/11/14-REGION 11 ANSWER 4.16 (2.50) '
l
-[Any FOUR (4) of the following FIVE (5), situations -
each worth 0.25 points]
c) 1) Entry into areas posted SPECIAL RADIATION WORK 1 PERMIT REQUIRED FOR ENTRY.
- 2) Entry into a reactor containment building. l
- 3) Job assignments where whole body exposure to any individual is likely to exceed 100 mram per entry.
- 4) . Job. assignments involving activities that have the potential for significantly increasing radiation or contamination - i.e.; opening primary sys boundaries.
- 5) Entry into " Exclusion" Areas. ;
b) Routine Radiation Work Permits are written for those creas where radiological hazards are stable or predictable and routine entries are necessary for plant operation. (1.0) c) TRUE (0.5)
REFERENCE l
'Fcr1cy Health Physics Manual, FNP-0-M-OO1 AND Qual Req'mt Adnin Procedure Area #13 KAIR 2.8/3.4 194001K103 ...(KA*S)
s
- 4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY 4tH), PAGE X123 o
RADIOLOGICAL CONTROL ANSWERS -- FARLEY -88/11/14-REGION II i
)
ANSWER 4.17 (1.00) i 1
"In cases of emergency... (0.1) I all personnel are authorized to depart {
from plant procedures provided that such l I
departure is necessary to prevent...
injury to personnel... (0.3) danger to the public ... (0.3) or damage to the facility." (0.3)
REFERENCE Fceloy Administrative Procedure FNP-0-AP-6, " Procedure Adhsrence" AND FNP-0-AP-16, " Conduct of Operations -
Oparations Group", Section 5.0, " Plant Operating Procedures" AND (Antonym / Contrast) Operations Memorandum 84-06 KAIR 2.5/3.4 4.1/3.9 l 194001A103 194001A102 ...(KA*S) l I ANSWER 4.18 (2.00) i NOTE: The candidate must express - either by assumption or by showing in the calculations the following:
To exceed 2200 mrem /qtr W.B. (which is the maximum the individual can go to before they must get the General 1 Manager's permission to exceed) the individual would j need to get another 1450 mrem, therefore..;
1450
= 29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br /> would be the maximum allowable time (0.5) 50 CALCULATION or ASSUMPTION worth 1.5 NOTE: If the calculation expresses *1450 the aspect of HP Manager's limitations has been presented - give credit j for this item.
REFERENCE Ferloy .LP OPS-310, Objectives 11 and 12; HP Manual FNP-0-M-OO1 AND Qual Req'mts Administrative Area l KAIR 2.8/3.4 194001K103 ...(KA*S)
I
O MASTER
- - 4 COPY ,
U. S. NUCLEAR REGULATORY COMMISSION.
,5 SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: FARLEY 1&2 Li REACTOR TYPE: PWR-WEC3 !
h DATE ADMINISTERED:- 88/11/14 !
EXAMINER: ' REGION II <
CANDIDATE:
INSTRUCTIONS TO CANDIDATE: .]
f Use separate paper for the answers. Write answers on one side only.
g Staple. question sheet- on top of the answer sheets, Points for each-t<
question are indicated in parentheses after the question. The passing j grade requires at least.70%.in each category and a final grade of at i
'least 80%. Examination papers will be. picked up six (6)- hours after the examination starts. l j
i
% OF' CATEGORY % 0F CANDIDATE'S CATEGORY VALUE' TOTAL SCORE VALUE CATEGORY 30.00 -25.00 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 30.00 25.00 6.
PLANT SYSTEMS DESIGN, CONTROL, j AND INSTRUMENTATION l 30.00 25.00 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL __ ,
_30.00 25.00 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 120.00 % Totals Final Grade All work done on this examination is my own. I have neither given e nor received aid.
i Candidate's Signature j 1
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I 5; I IHEOI4Y OF ' NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS PAGE 2.
l
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p sys W .bbt QUESTION 5.01 '(1.00)
'0PY The primary reason delayed neutrons make the reactor more controllable is-dolayed neutrons have: [ SELECT THE CORRECT REASON]-
A. A shorter life time 1 B. A longer life time C. Higher energy !
D. Lower' energy l
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3 i THERMODYNAMICS MASTER QUESTION 5.02 (1.00)
.Whether the moderator. temperature coefficient is positive or negative
.dopends on: [ SELECT THE CORRECT STATEMENT]
A. Whether the given water-boron mixture absorbs more neutrons than it' moderates or moderates more neutrons than it absorbs B. Only the boron concentration because raising the boron ;
concentration makes the moderator temperature coefficient more 'l negative.
C. Only the moderator temperature because the magnitude of the moderator temperature coefficient is greater at higher temperatures D. Both the boron concentration and moderator temperature because although both the moderator pressure and void coefficients are positive they are neglected
[
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i
f a i C. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMIC'CS PAGE 4 MAS.*it*il QUESTION 5.03 (2.00)
COPY At 0001 on November 1, 1988 a reactor shut down from 100 percent power was initiated. Over the next 3 days the reactor was shutdown and cooled down to 140 degrees F. During the cooldown, boron concentration was increased by 100 ppm. Given the following absolute values of reactivity change over the 3 days:
Xenon = [.] 2575 pcw Rods = [ ] 6938 pcm Temperature = [ ] 550 pcm Boron = [ ] 1140 pcm Power Defect = [ ] 1525 pcm A. What was the shutdown margin at 0001 on November 1, 1988? [ SELECT THE CORRECT VALUE]
- a. - 5413 pcm
- b. 5963 pcm
- c. - 6938 pcm
- d. - 7913 pcm B. What is the chutdown margin after all the reactivities were added?
[ SELECT THE CORRECT VALUE]
- a. - 3428 pcm
- b. - 4863 pcm
- c. - 6478 pcm
- d. - 9053 pcm l
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) 5.
' THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 5 THERMODYNAMICS
=p.
1 eg'py QUESTION 5.04 (1. 00)
Aft.cr operating at 50% power for several days, reactor power is increased to 100%. How will reactivity change in the next one to two hours? [ SELECT THE CORRECT ANSWER]
! A. Positive due to the faster burnup of Xenon at higher neutron flux levels B. Positive due to the incraased rate of decay of Xenon to Cesium C. I ga ive due to the increase production of Xenon at the higher D. Constant initially due to the 6.2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> half-life of Iodinc 135
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QUESTION 5.05 COPY (1.00).
A reactor has been' operated at 100% power for.100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. Which'one of the following statements best describes Xenon behavior over the first few hours following'.a power decrease?
.[ SELECT THE CORRECT ANSWER] I NOTE: [Xe] denotes xenon concentration A., Direct (Xe] increases, indirect (Xe] decreases, total [Xe]
decreases.
B.. Direct'[Xe] increases, indirect [Xe] increases, total [Xe]
increases. !
C. Direct,[Xe] decreases, indirect (Xe] decreases, total [Xe]
decreases.-
.Direct [Xe decreases, indirect [Xe] increases, total [Xe]
D.
increases.]
E.' Direct decreases.
[Xe] dpcreases, indirect [Xe] increases, total [Xe]
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'5. THEORY OF NUCLEAR POWER PLANT /0PERATION, FLUIDS, AND THERMODYNAMICS PAGE 7 A
[k h fiT"P t, }% '
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-QUESTION 5.06 (1.00) JJ w. - '
l Why are rod insertion limits specified to maintain a minimum rod height during reactor operations? [ SELECT THE CORRECT ANSWER]
A. Maintaining the rods above the rod insertion limit minimizes the ~
effects of Xenon on radial and axial flux difference B. Maintal'ning the rods above the rod insertion limit minimizes the required shutdown margin C. Maintaining the rods above the rod insertion limit ensures that rod position indication misalignment is not exceeded ;
i D. Maintaining the rods above the rod insertion limit ensures that 1 acceptable power limits are maintained l
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- 5. THEORhOFNUCLEARPOWERPLANTOPERATION,. FLUIDS, AND PAGE 8 THERMODYNAMICS p c.n }
fih): msk QUESTION 5.07 (1.00)
A reactor startup ic in progress. The source range count rate indication was 11 cps when the startup commenced and now indicates 2386 cps. What is the expected count rate response resulting from a brief rod withdrawal-during the approach to criticality? (SELECT THE CORRECT COUNT RATE RESPONSE]
A. An immediate rapid rise continuing to criticality B. An'immediate rapid rise followed by a gradual increase to a higher steady state value C. A gradual increase followed by a rapid decrease when rod' withdrawal is stopped D. A gradual increase continuing to criticality 1
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i
5.- THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 9
' THERMODYNAMICS MASTER QUESTION 5.08 (1.00) COPY j i
During a reactor startup, control rods are withdrawn such that the new l
- . steady-state count rate is 800 cps. If the count rate before rod '
withdrawal was 400. cps and Keff was 0.96, what is Keff following rod withdrawal? (SELECT THE CORRECT Keff] j A. 0.97 B. 0.975 C. 0.98 D. 0.985 1
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.5;:hHEOR'YOFNUCLEAR~POWERPLANTOPERATION, FLUIDS, AND' PAGE'10.
THERMODYNAMICS -
QUESTION 5.09 (1.00)
If reactor power is increased at a constant 1.0 dpm startup rate up to the.
point of adding heat,. which one of the following responses will be the INITIAL indication that the point of adding heatthas been' reached?-(SELECT THE CORRECT INITIAL INDICATION OF REACHING THE POINT OF ADDING HEAT]
A. RCS temperature increases B. Pressurizer level increases C.. Reactor power stabilizes D. Startup rate decreases
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 11 l THERMODYNAMICS . 7' h O QUESTION 5.10 (2.00)
COPY For each of the following events, state whether the differential rod worth l of an individual control rod will INCREASE, DECREASE or REMAIN THE SAME.
Consider each case separately.
A. An adjacent rod is inserted to the same height.
B. Moderator temperature is increased.
C. Boron concentration is decreased.
D. Fuel adjacent to the control rod is depleted.
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~ .S . THEORY OF NUCLEAR POWER PLANT OPERATION, FLU PAGE 12
- THERMODYNAMICS COPY QUESTION 5.11 (2.00)
A critical' boron calculation has been performed prior to startup. State how the calculated value changes for each of the following. Answer INCREASE DECREASE or REMAIN THE SAME.
A. The REFERENCE critical condition for bank D rods is corrected from 160 steps withdrawn to 180 steps withdrawn.
B. The DESIRED critical rod height is increased from 160 steps withdrawn to 180 steps withdrawn.
C. The REFERENCE critical condition for reactor power is corrected from 80 percent reactor power to 75 percent reactor power.
D. Estimated time of startup is shortened from 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after shutdown from 100% power. l
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- 5. l THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 13 THERMODYNAMICS {
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QUESTION 5.12 COPY (1.50)
Answer TRUE or FALSE for each of the following statements concerning the count rate (inverse multiplication) (1/M) plot for rod withdrawal .
A.
As the reactor approaches criticality, the critical rod height prediction method becomes more accurate. 1 I
B.
Variation in the magnitude of differential rod worth has no effect on the critical rod height prediction.
C.
A count rate, which is taken before the reactor power level reaches steady state (i.e. count rate is taken shortly after reactivity is added), will result in a HIGHER predicted critical rod height than if a steady state count rate were taken.
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 14 -1 THERMODYNAMICS j
MASTER l QUESTION 5.13 (2.00) ;
Two identical reactors are taken critical. Reactor A has a rod speed of 40 steps per minute. Reactor B has a rod speed of 30 steps per minute. Assuming a continuous rod withdrawal in each case, answer: A, B, or THE SAME to each of the following questions. ,
A. Which reactor will achieve criticality first?
B. Which reactor will have the highest source range count at criticality?
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND
. THERMODYNAMICS PAGE'15 MAS 9h. R QUESTION 5.14- (1.00)
CDPY !
-Ilow will pressurizer level indication compare to actual pressurizer level oif containment temperature is increased [cteam leak] from 100 degrees F to l
180 degrees F? [ Consider only reference leg heating effects] [ SELECT THE CORRECT COMPARISON] l i
A. Indicated pressurizer. level will be higher than actual !
i pressurizer level B. Indicated pressurizer level will be lower than actual pressurizer level i
l C. Indicated pressurizer. level will be equal to actual pressurizer level D.
Indicated. pressurizer level will fluctuate as a result of heating the reference leg i
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5e -THEORY OF' NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS- PAGE 16 MASTER QUESTION 5.15 (1.00)
Unit leak.
1 is operating at 30% load following a 2 day outage to repair a steam compensation A ramp failed increase in load at the 30%isload commenced with the steam flow density value. How will the 100% load steam-flow indication be affected by the steam flow density compensation failure?
(SELECT THE CORRECT COMPARISON)
A. The indicated steam flow will be greater than actual steam flow B.
The indicated steam flow will be less than actual steam flow C. The indicated steam flow will be equal to actual steam :flov l D.
The indicated steam flow will fail at the 30% load value i I
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- 5. h'HEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND , , PAGE 17 THERMODYNAMICS Q- .;
B(9[1E D d .Sh) .
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COPY l QUESTION 5.16 (1.00) l
?
Which of the following' steam generator parameters.is nearly linear l
[directly proportional) to reactor power as turbine load is ramped from 25%- j to 75% load? [ Rod control is in automatic] [dELECT THE CORRECT PARAMETER]
A. Delta-h [ steam enthalpy - feed enthalpy)
B. Delta-T [ steam temperature - feed temperature]
C .' Steam flow D. Steam generator level l
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- 5. THEORY OF' NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS PAGE.18 !
' QUESTION 5.17 (1.00) 1 Which of the following indications DOES NOT' verify that natural circulation
- is established? '[ SELECT THE INCORRECT INDICATION] j j
A. Core exit thermocouple . stable or trending down B. Decrease'in steam pressure during constant decay heat j C. Steam generator pressures stable or trending down !
l D. RCS cold leg temperatures at saturation temperature for steam generator presssure l
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l L5.
THEORY OF NUCLEAR POWER PLANT: OPERATION, FLUIDS, AND PAGE 19 THERMODYNAMICS'
? Nill l sir
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' QUESTION- 5.18 COPY ;
(1. 00) I The following are present plant-conditions in Unit 2:
- Mode 3
- MSIVs shut
- Primary system and S/Gs in thermal' equilibrium
- S/G pressures are 915 psig t
- PZR pressure is 1615 psig '
Which statement most correctly describes the subcooling margin (SCM) for the-given conditions?
[ SELECT THE MOST CORRECT STATEMENT]
A. Present SCM is less than the expected SCM at normal 100% rated thermal power (RTP) operation.
B. Present SCM is the same as the expected SCM at normal 100% RTP operation.
C. Present SOM is greater than the. expected SCM at normal 100% RTP operation.
(L L D. The relative: magnitude between the present SCM and the expected SCN j at normal 100% RTP operation can not be determined due to insufficient information given to calculate the present SCM., ,
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..5. : THEORY OF NUCLEAR POWER PLANT OPERATION,- FLUIDS,-AND PAGE 20 '
THERMODYNAMICS M. STER l QUESTION 5.19 ( 1. 00 )'
A. pipe'with.superheated steam at 240 psia and 1000 degrees F develops a small' crack causing an iscenthalpic expansion of..the steam to atmospheric pressure. What'is the condition of the fluid outside-of the pipe? [ SELECT THE MOST CORRECT STATEMENT]
NOTE: Use steam. tables only; do not use Mollier Diagram.. l l
A. Fluid remains a superheated vapor with greater than 100 degrees F j of superheat. l 1
B. Fluid remains a superheated vapor with less than 100 degrees F of R superheat.- !
C. Fluid becomes a saturated vapor with greater than 10%' moisture
. content.
D. Fluid becomes a saturated vapor with less than 10% moisture-content.
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I 5 ~. THEORY OF NUCLEAR POWER PLANT OPEP.ATION, FLUIDS, AND PAGE 21-THERMODYNAMICS MASTER QUESTION 5.20 (2.00) 00PY Unit.1 is operating at 30 percent power with rod control in manual.
For each of the following parameters, state whether an inadvertent closure of S/G A main steam isolation valve would cause the parameter to INCREASE, DECREASE or REMAIN THE SAME. Consider each' case separately. Assume power is maintained at 30 percent.
A. Departure from Nucleate Boiling Ratio (DNBR).
B. Reactor Coolant System Sub' cooling Margin (RCS SCM).
C. RCS Loop B Hot Leg Temperature. i D. Quadrant Power Tilt Ratio (QPTR). ,
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'5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS,,AND PAGE 22 THERMODYNAMICS ..*
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. QUESTION. 5.21 (1.00)
. TRUE or PALSE' for each of the following:
A. During an RCS heatup, as' temperature gets higher,'it will take a sm'ller a letdowri flow rate-to maintain a constant pressurizer 3evel.
B. Increasing condensate depression (subcooling) will cause BOTH a {
decrease in plant efficiency AND an increase in condensate ;
(hotwell) pump available NPSH. '
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS PAGE 23 ul/ '
QUESTION 5.22 (2.00) 31PY l
For each of the following, indicate whether the Departure from j I
Nucleate Boiling Ratio will INCREASE, DECREASE or REMAIN THE SAME.
Consider each case separately. l A. One reactor' coolant pump trips resulting in two loop power operation.
B. Reactor power decreases.
C. One rod drops D. Automatic pressurizer spray initiates.
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g+ . .
?! 5. ' THEORY OF NUCLEAR POWER' PLANT OPERATION, FLUIDS, AND
' PAGE 24 THERMODYNAMICS MASTER
. QUESTION 5.23 (1.50)
{N.
4
= Assume that the power range channels have been adjusted based.on a-I - calculated calorimetric. Answer each'of the following statements'TRUE 3 l
ar FALSE.
]
A. 'If the blowdown flow had been ignored in calculating the calorimetric, k a
then actualoreactor power would be higher than indicated reactor
' {
power.
D. If the feedwater temperature used in calculating the calorimetric had i been 10 degrees lower than actual feedwater temperature, then actual reactor power would be higher than indicated. reactor power.
C. If a main steam atmospheric relief valve had been leaking by during l the data-taking portion'of the calorimetric, then actual reactor power would be. higher than.ir.dicated reactor power.
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(***** END OF CATEGORY 05 *****)
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. 6 2- PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION- PAGE 25 l
k j
. QUESTION .6.01 (2.00)
COPY '
- The rod' control reset switch on the MCB perfonns six functions, including: -)
1.- Resets the master cycler 0-5 counter to zero,
- 2. Resets the slave cyclers 0-127 counter.to zero, l
- 3. Resets the bank overlap unit 0-999 counter to zero, and
. Resets the pulse-to-analog convertors in the rod position 4.
indication system to zero.
A. What are the other functions performed by the. rod control reset switch?' [ LIST:TWO) .[1.0] f 1
i B..
. Which of the functions ensures correct insertion' limit surveillar.ce?
[ SELECT FROM FUNCTIONS 1-4 LISTED ABOVE) {0.5]
'C. Which of the functions ensures correct sequence of a step? [S EL"ECT.
FROM FUNCTIONS 1-4 LISTED ABOVE) [0.5]
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,, .6. PLANT SYSTEMS DECIGN, CONTROL, AND INSTRUMENTATION' PAGE 26' l
. I ; ) Q ?* y
%A& b -fil%
. QUESTION 6.02 (2.00)
A. Why is the actual. rod position indication a more reliable indicator of rod. position'than the demand red position indication? (1.5] .i
, B .~
What are the systems that provide domand and actual rod position irsdication? [ STATE THE SYSTEM FOR EACII INDICATION] (0.5]
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?
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.3 . .
- 6. PLANT' SYSTEMS-DESIGN, CONTROL, AND' INSTRUMENTATION! '
PAGE 27 MASTER 1 COPY 1 QUESTION 6'03 (2.00)
A. .What are the TWO [2] sources of water for the. containment spray 1punips? Ij 1[ LIST TWO (2)]'[0.4] .L
^)
.?
B. Whatare the' conditions, including setpoints and' coincidence, that . j will. generate a."P" signal?'[0.4] 1 1
C. Stat'e .the. THREE. [3]. containment apray ' system coraponents which reposition / start'in. response to a "P"' signal?' Include in yourLanswer jl the' component's final position et state. Parallel.or similar items '
count as one componeht. fl.2]
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'6.', PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 28 MASTER ;
-QUESTION 6.04 (2.50)
COPY REFER TO THE ATTACHED DRAWINGS OF THE CONDENSATE SYSTEM AND MAIN FEED SYSTEM WHEN ANSWERING THE FOLLOWING QUESTIONS:
g A. How is the A S/G feed pump discharge valve, V503A, interlocked? [0.5]
B. What.is the purpose of the component designated LEFM?_[0.5]
C. What is the purpose of the motor operated check valves MOV-3232A, B, and C? [0.5]
D. What is the purpose of V904? [ INCLUDE SETPOINT) [0.5]
E. How will V902 fail on a loss of air? [0.25]
F. What signal controls the position of V909B? [ IDENTIFY THE COMPONENT ORIGINATING THE SIGNAL) [0.25] I l
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(***** CATFGORY 06 CONTTNUED ON NEXT PAGE *****)
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6 '. . PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 29 L
MASTER l QUESTION 6.05 (3.00) nnw iJUjl 3
-What are the interlocks _and/or control actuations for the following Chemical And Volume Control System (CVCS). components: [ INCLUDE SETPOINTS AND FAILURE MODES AS APPROPRIATE] l i A.- Letdown-Isolation Valves (LCV-459, 460)?
I B. Letdown Orifice Isolation Valves (8149A, B, C) ?
C.
l.
! VCT Level Control Valve (LCV-115A)?
D. VCT Outlet Isolation Valves (LCV-115 C, E)?
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- 6. PLANT SYSTEMS DESIGN, CONTROL, AND. INSTRUMENTATION FAGE 30
)
MASTER f
QUESTION. 6.06 (2.70) C.we u l
-What'are pumps:
the automatic start signals for the following auxiliary-feedwater
[ INCLUDE SETPOINTS AND COINCIDENCES AS. APPROPRIATE]
A. Motor D: liven Auxiliary Feedwater [MDAFW) Pump?
B.- Turbine Driven Auxiliary-Feedwater (TDAFW) Pump?
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_-_-__-- - _ - . . _ - _ _ _ . _ i
I 6.J: PLANT SYSTEMS DESIGN,-CONTROL, AND' INSTRUMENTATION PAGE 31-ESTER d
fQUESTION 6.07 COPY 1 (2.50) .-
Thaimanipulator crane is. equipped with various-interlocks designed-to-prevent damage.to a fuel assembly. .-
A. What prevents aLfuel' assembly'from.being picked up' partially-engaged?-[0.5]
B. What is.the: purposeof the overload interlocks? [0.5).
.C. What does the slack cable-interlock prevent? (0.5]
'LD.- What-do the slow zones ensure?-(1.0]
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- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 32 MASTER )
QUESTION 6.08 (2.50) 03PY l
)
i If both 48V and 15V power supplies have been verified operational, what f additional faults or actions would be checked to determine the cause of a Train "A" Solid State Protection System (SSPS) general warning alarm?
[ LIST FIVE (5)] :
]
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- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION JPAGE 33' MASTER QUESTION 6.09 (2.50)
COPY A. What~are the electrical sources of fault signals that will initiate' '
fast dead bus transfer of'4160V buses A, B, and C? [ LIST THREE (3)
{
DIFFERENT SOURCES) lB . .-
Following a fast dead bus transfer, what is.the power supply for 4160V buses A, B, and C?
i
[ STATE THE ALTERNATE POWER SUPPLY FOR THE BUSES]
C.-
If the fast dead bus transfer is not completed within 15 cycles, why- .!
is the fast dead bus transfer blocked? [ EXPLAIN WHY THE. TRANSFER I WOULD NOT BE NECESSARY) i 1
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- 6. ,PLANTLSYSTEMS DESIGN, CONTRO.L, AND INSTRUMENTATION PAGE 34 Lt t jfVh .
9IP '
d d (h}l QUESTION ~6.10 (1.80) ~
~
. Refer to the attached drawing of the Reactor Coolant System (RCS) to identify the following RCS penetrations: [ IDENTIFY /NAME OR STATE THE
. PURPOSE OF PENETRATION AND INDICATE TO/FROM AS APPROPRIATE]
'A. Penetration number 3?
B. Penetration number 4?
C. Penetration number 10?
D. Penetration number 11? [ question deleted) i E. Penetration number 12? ?
F. Penetration number 14? [ question deleted) !
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M A'N C00_A C .00 ? 3 E.E~1A" Ob S MASTER C]PY t
- ,._____________--,__________________,I NOTE 1
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',', NOTE 1 ,# 15 PUMP i' 8 n 14 3 6 14 i'
- 7 i
l PUMP g2 16 5,'
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- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 35 l
MASTER QUESTION 6.11 (2.00)
Refer to the attached-Pressurizer Pressure Protection And Control diagram j to answer the following if the master pressure controller fails at 43%
[10'32 ma] while selected to automatic: [ ANSWER TRUE OR FALSE] [ ASSUME NO OPERATOR ACTIO..]~
i A. Variable heater group C voltage will decrease to cause pressure.to ;
decrease, i B. Spray Valve PCV 444C will open to maintain pressure within the operating band.
C. PORV 444B will open when pressure increases to setpoint.
D. PORV 445A will open when pressure increases to setpoint.
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i
6 -
COPY .
'l CONTROL PROTECTION CONTROL
@ @ u ::,
3 I - E 4 (IDENTICAL TO Il ll0ENTICAL TO Il MCB MCB MCB
- PI- - 1/A -- Pl . - PI
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RESET BLOCR RATE l
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- ISOLATION VALVES
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.SAftTY R TRIP R TRIP INlECil0N 4
pg l RECORDER l -
ri iNC 414 * ,,. ..
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'l li it l' f 8ACKUP . ALARM V ARI ABLE AL ARM $ PRAY $ PRAY PORV PORY ALARM ALARM l MtAT(R BACKUP MEATER CONT VALVI VALVE 444B 445A PZR PZR GROUP HTR$ GROUP OUTPUT PCV PCV PRESS LO A801 ON C HIGH 444C 4440 Hi PRES $
PRESSURIZER PRESSURE PROTECTION AND CONTROL
. FIGURE 3 . . .
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- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 36 h1 ASTER fh -
QUESTION 6.12 (3.00) S8 w e E A. What are the purposes of interlocks associated with the RHR inlet isolation valves? [ DESCRIBE / DISCUSS THREE (3)]
B. What are the interlock requirements to open RHR inlet. isolation valve 8701B? [ DESCRIBE / EXPLAIN THREE (3) AND INCLUDE SETPOINTS AS APPROPRIATE]
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- 56. PLANT SYSTEMS DESIGN, CONTROL, AND-INSTRUMENTATION PAGE 37 l
MASTER
' QUESTION- 6.13 (0.50) 1 The' Component Cooling Water (CCW) System Surge Tank is divided into TWO sections by a metal partition that extends from the bottom of the tank to(2]
about three-fourths of the height of the tank.
What is the purpose of this tank separation?
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-l .6 . - PLANT. SYSTEMS DESIGN,. CONTROL, AND INSTRUMENTATION PAGE 38 ,
1 M,ASu..,..R e L
1 00PY
-QUESTION 6.14 (1.00) i --
l By design,.the maximum actual, capacity of any single Main Steam Atmospheric '
Relief Valve, at a steam pressure of 1085 psig will'not exceed 890,000lb/hr. '
What'is-the reason for thisllimitation?-
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L __ __ __ _ __ ___
.7 s - PROCEbURES-NORMAL, ABNORMAL,. EMERGENCY AND PAGE 39 L
RADIOLOGICAL CONTROL
?$
MAS"I bk r,
QUESTION 7.01 (2.25) E Unit 1 is operating at 87% load'when a PRESSURIZER PRESSURE HI-LO ALARM is received. Following the alarm, the operators report:
Pressurizer pressure 2300 psig and decreasing What are the AOP-17.0, Malfunction Of RCS Pressure Control System,
'immediate operator actions that should be completed or performed?
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s.
'.7.. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND ~PAGE 40-RADIOLOGICAL CONTROL y i
CDPY QUESTION ~7.02 (2.00)
- Refer to the attached SOP 28.1, Turbine Generator Operation, Figures 1, 2, and 3:
A. 'When' perfor:ning a cold turbine, startup, .what acceleration rate should be used to accelerate to synchronous speed? [ ASSUME INITIAL HP TURBINE FIRST-STAGE METAL TEMPERATURE OF 100F]
B.: 'How much time is required to change from 10% to 50% load? [USE A FATIGUE INDEX = 20000 CYCLES]
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7, PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 41 RADIOLOGICAL CONTRCL MASTER QUESTION 7.03 (1.50)
The Unit one operators report rapidly decreasing condenser vacuum. What are the AOP 8.0, Partial Loss Of Condenser Vacuum, immediate operator actions? [ LIST THREE (3)]
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- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND - PAGE 42 RADIOLOGICAL ~ CONTROL MASTER
' QUESTION 7.04. (1,00)
An operator has~a whole body radiation exposure of 150 mr for'the current quarter.
How long can the operator stay in a 50 mr/hr radiation area without exceeding the maximum quarterly whole body dose limit that may be'
. authorized by the HP. manager? [ SELECT THE CORRECT ANSWER]
A. 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> B. 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> C. 41 hours4.74537e-4 days <br />0.0114 hours <br />6.779101e-5 weeks <br />1.56005e-5 months <br /> ;
D. 57 hours6.597222e-4 days <br />0.0158 hours <br />9.424603e-5 weeks <br />2.16885e-5 months <br /> i
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- 7. PROCEDUPES - NORMAL, ABNORMAL, EMERGENCY AND I PAGE 43 l RADIOLOGICAL CONTROL l
MASTER )
QUESTION 7.05 (2.00)
,,py JJ Answer TRUE or FALSE for each of the following statements.
A. The NRC external dose limit to the skin of the whole body is 18.75 rem B. The NRC quarterly whole body limits are 3 and 12 rem C. The Emergency Director is the only person authorized to approve emergency doses D.
The year senior vice president's approval is required to exceed 5' rem per i
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l 7., PROCEDURES - NORMAL, ABNOR!iAL, EMERGENCY AND j
RADIOLOGICAL 1., CONTROL PAGE 44 MA3iid ,
QUESTION 7.06 (1.50) P"E3PY !
j
[ ANSWER TRUE OR FALSE FOR EACH STATEMENT]
Which of the following statements are accurate precautions or limitations for operation at pot..ar in accordance with FNP-1-UOP-3.1, POWER OPERATION?
A. When the generator is on the line, then do not adjust generator excitation with the manual voltage adjust switch.
B. If outside 20.9 - 23.1 kV when adjusting VAR's, then reduce generator output due to stator overheating.
C. The administrative limit for main generator MVAR's is -300 to prevent f stator overheating.
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- CATEGORY 07 CONTINUED ON NEXT PAGE * * * * *)
.7.- PROCEDURES ' -' NORMAL, ABNORMAL; EMERGENCY AND RADIOLOGICAL CONTROL PAGE 45 h r" .1 @
bb $'Ad) d (
QUESTION 7.07- (2. 00) . ,
Following a reactor trip on' Unit'1 the operators have properly exited.to '
l FNP-1-ECP-0.0, . Loss Of.All AC Power, and completed immediate -action steps
.1 & 2 ... How does the operator perform immediate action Step 3. " check RCS isolated"? .)
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-[ LIST.THREE STEPS INCLUDING VALVES AND SETPOINTS AS APPROPRIATE] '
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- 7. : :' PROCEDURES ~ - - NORMAU, ABNORMAL, < EMERGENCF IND - PAGE 46'
. RADIOLOGICAL CONTROL x - MASu,..,d f
QUESTION 7.08 (1.00) l Following a Unit i reactor trip caused by a. spurious turbine trip, operators have properly exited to FNP-1-ESP-0.1, Reactor Trip Response, and have completed Step 7.2..
[ SELECTED PAGES FROM FNP-1-ESP-0.1.ARE ATTACHED]
The following conditions currently exist:
Tavg trending to - 547F RCG pressure ~.1855 psig Pressurizer levelL~ 20%
Nuclear Power - 0%'
, SG level - 40% inall SG CTMT pressure ~ 5 psig ,
Core exit TCs - 600F i
As the Shift Supervisor, what action will you direct the operators to take.
in accordance with FNP-1-ESP-0.1?
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= - - - _ _ _ - _ _ _ _ _ _ - _ _ _ _ -- _ -
-]
= ,
j 7.: . PROChDURES L- NORMAL, ABNORMAL, EMERGENCY AND PAGE 47 RADIOLOGICAL CONTROL 3 p ,.
dks)t l !
-QUESTION 7.09 (2.75)
COPY i
The.following pertains to FNP-1-FRP-S.1, Response to Nuclear Power j Generation /ATWT, immediate action step 4., Determine if. emergency boration !
required.
l A. Complete the following statement: [ FILL IN THE BLANKS]
Borate (1) ppm for one rod and (2) ppm additional I for each additional rod NOT full inserted. [0.5]
B. What action would you perform /take/ direct if immediate action step 4.2, Start at least one boric acid transfer pump (on service pump preferable), can not be successfully completed? [ STATE THE RESPONSE NOT OBTAINED ACTION AND INCLUDE THE COMPONENTS TO BE OPERATED] (1.0]
C. What action would you perform /take/ direct if immediate action step l 4.3, Open emergency borate valve, can not be successfully completed? l
[ STATE THE RESPONSE NOT OBTAINED ACTION AND INCLUDE THE COMPONENTS TO >
BE OPERATED) [1.25]
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- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND l RADIOLOGICAL CONTROL FAGE 48 i MASTER QUESTION 7.10 (3.00) ) )[
EIPs-12, 17, 18, & 19 ARE ATTACHED:
A. Refueling operations are in progress. A fuel assembly has been fully withdrawn from the core. When the operator starts to move the i manipulator crane towards the upender, the manipulator crane fails.
The operators can neither lower the fuel assembly back into the core )
nor move it to the refueling canal. Immediately after the manipulator crane failure the refueling cavity water level rapidly decreases to the top of the core.
What is the emergency classification [ NOTIFICATION OF UNUSUAL EVENT, ALERT, SITE AREA;;EMERGENCY OR GENERAL EMERGENCY] for the given plant conditions? ,.
B. At 0925 a private aircraft crashed onsite. Approximately 10 seconds after the crash a loss of off site power occurs. ECCS actuation and i high secondary collant activity occurs.
What is the emergency classification [ NOTIFICATION OF UNUSUAL EVENT, ALERT, SITE AREA EMERGENCY OR GENERAL EMERGENCY] for the given plant 3 conditions?
C. At 0924 power was removed from all main control board annunciators to isolate a ground. At 0925 a private aircraft crashed onsite.
Approximately 10 seconds after the crash a loss of off site power occurs. The operators properly execute the Emergency Event Procedures and Emergenc) Contingency Procedures. At 0935 the aircraft pilot has been removed from the aircraft. At 0945 main control board annunciators are restored. At 0946 the injured pilot is trailsported to the hospital by ambulance.
What is the emergency classification [ NOTIFICATION OF UNUSUAL EVENT,
! ALERT, SITE AREA EMERGENCY OR GENERAL EMERGENCY) for the given plant
[ conditions? -
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, 7.. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 49 RADIOLOGICAL CONTROL ,
1 QUESTION 7.11 COPY I (3.00)
A reactor trip has occurred and the operators are performing FNP-1-EEP-0, Reactor Trip or. Safety Injection. The operators have progressed to immediate action Step 4. Check if SI actuated.
[SEE ATTACHED PAGE FROM FNP-1-EEP-0]
A. How do the operators perform Action / Expected Response Step 4.1 "Any SI actuated indication"? [ LIST TWO DIFFERENT INDICATIONS] [0.9]
! . lB .
How do the operators perform Response NOT obtained Step 4.1.1, "Tr'ip status light box bistables meet coincidence for SI", and Step 4.1.2,
" Parameter indicators have reached an SI setpoint"? [ LIST FIVE (5)
INCLUDING SETPOINTS AND COINCIDENCES AS APPROPRIATE] [2.1]
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-FNP-1-EEP-0 00P Reactor Trip or , fe Injection. Revision 8 O' -Step Action / Expected Response Response NOT Obtained' .
L l l l l l l 4 4 4 l t
4 4 4 .
I 2 Check turbine tricted. 2 Perform the following.. l TSLB2 14-1 lit ~ j l: -
TSLB2 14-2 lit' ~
TSLB2 14-3 lit 1
[TSLB214-4 lit 2.1 Manually trip turbine. 'l l
MN WRB 1 EMERG TRIP I
-to TRIP. position and hold '
for approximately 5 seconds. j 2.2 IF turbine still NCTr trip l
-THEN stop EH fluiT ~ pumps. ped,
]
EH FLUID '[
PMP .l
_1A off -)
1B off .
3 . Verify at least one train of 3 IF both trains deenergized, 4160 ESF busses energized THEN go to FNP-1-ECP-0.0, ;
(F & K or G & L white cower LOSS OF ALL AC PCWER. <
available lights lit).
4 Check if SI actuated j
. l 4.1 Any SI actuated indication 4 '.1 IF either of the following !
present, THEN manually -
J.s actuate SI and continue with l'.
Step 5. !
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'4.1.1 Trip status light box bistables meet coincidence for SI.
4.1.2 Parameter indicators have i
reached an SI setpoint.
4.2 IF SI NCTr required, THEi go
~
I TRIP RESPCNSE.
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- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND l PAGE 50 <
RADIOLOGICAL CONTROL 4 7Mi bgglER \
QUESTION 7.12 (3.00)
COPY !
A. . As the Reactor Operator, what' action would you take in accordance with ;
AOP-19.0, Malfunction of Rod Control System, if an unexplained outward 1 rod motion resulting in a significant positive reactivity addition occurred? [ LIST TWO ACTIONS] (1.0]
B. Given the following indications:
One [1] Rod Bottom Light Rod at bottom. alarm Sudden drop in Tavg 1.
What are the automatic actions that should occur? [ LIST TWO]-
(ASSUME AUTO' ROD CONTROL] (1.0] ;
- 2. What immediate operator actions should be taken in accordance with FNP-1-AOP-19.0, MALFUNCTION OF ROD CONTROL SYSTEM? [ LIST TWO. ACTIONS] (ASSUME AUTO ROD CONTROL] [1.0]
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- 7. PROCEDURES.--NORMAL, ABNORMAL, EMERGENCY AND "
' RADIOLOGICAL CONTROL' PAGE 51' Y
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-QUESTION -7.13' (1.00) I
.;According to FNP-O-AP-6, " Procedure Adherence": '
" Emergency deviations.from TECHNICAL SPECIFICATIONS are authorized-subject to-[THREE (3) "specified"] conditions.... ,
i LIST TWO -[2] of these THREE [3]'" conditions"-
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J '7o PROCEDURES '- NORMAL, . ABNORMAL, EMERGENCif AND
' PAGE 52 RADIOLOGICAL CONTROL MASTER QUESTION 7.14 (2.00) l()l'if A fireSupervisor Shift exists in thedecides Controlthat Room (around control of the centerfrom equipment deskthe section) control AND the room is in;joopardy.
. LIST FOUR (4) of FIVE (5) actions which should be performed - in an expeditious manner and in accordance with the " Fire In The Control Room" procedure (AOP-28.2) - both PRIOR TO and IMMEDIATELY AFTER the control i
, room evacuation and BEFORE assemblage of personnel at the Hot Shutdown
~ Panel. !
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- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 53 RADIOLOGICAL' CONTROL NAS.n.ER QUESTION 7.15 (2.00)
COPY LIST the FOUR (4) immediate operator actions of the Emergency Boration procedure FNP-1-AOP-27.0.
(***** END OF CATEGORY 07 *****)
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 54 7AASTER QUESTION 8.01 (1.50) 80PY' What THREE [3] actions should be taken in accordance with FNP-0-SOP-0, General Instructions To Operations Personnel, if log readings are not taken within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the specified time?
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~l "8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PARS 55 '
i NASTER'
'QUENTION 8.02 (1.00)
,0IPY Wh'ich of the fol}owing statements is correct regarding Locked Valves and Breaker. Key Controls?
A. For' operating locked valven in containmer,t and when personnel are '
in containment,.the Shift Supervicor may ha'e'a v key delivered to an individual at the containment entrance, B. For. operating locked valves in containment and when personnel are in containment, a Shift Foreman may.have a key delivered to an individual at the containment entrance.
C. The Shift Supervisor's key ring may be loaned only to Operations'
}
Group Supervisors or NRC inspectors.
o D. .The " supervision key ring", having master keys to doors will only ;
be checked out to Operations. Group Supervisors or NRC inspectors. {
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- 8. ADMINISTRATIVE PROCEDURES,. CONDITIONS, AND LIMITATIONS
-PAGE 56 e
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l EQUESTION' 8.03 (1. 0 0 )-
Which of the following statements is correct regarding manual operation of
.MOVs?
(SELECT THE CORRECT STATEMENT]
A.
DO NOT hold the declutch lever in the depressed position while the motor is running. This prevents motor overspeed damage.
B. DO NOT at any time depress the declutch lever. This prevents damage to the clutch internals.
C.
DO NOT use the manual operator to force the valve any further against its seat than the motor operator will drive it. The motor may not be able to drive the valve off the seat without damaging the operator.
D.
DO NOT use the manual operator to force the valve any further against its seat than the motor operator will drive it. The seat may be' damaged by the additional mechanical advantage of the mechanical operator.
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jc 8. c: ADMINISTRATIVE' PROCEDURES, CONDITIONS, AND. LIMITATIONS ~
i PAGE'57 i nL. MASTER f 1
-QUESTION L
i
- 8. iO 4 (1.00) 80PY
{ .- i.
How would you determine the.use/ purpose of a: key reported as found in the W turbinn building and numbered DE-243. -[ STATE THE PROCEDURE YOU WOULD CONSULT]
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8 .- ADMINISTRATIV
E. PROCEDURE
S, CONDITIONS, AND LIMITATIONS PAGE 58i 4
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MASTER QTESTION 8.05 (1.50) )
- What are the two conditions that EIP-9, RADIATION EXPOSURE ESTIMATION AND 'I (CLASSIFICATION OF EMERGENCIES, provides as the basis for classifying'an 3
-emergency? (LIST TWO). !i
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. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS ;
!- PAGE 59 L
g MASTER QUES: TION 8.06 (2,00)' '30PY Givsn the following plant conditions, what action (s) should be taken in accordance with technical specifications? [ INCLUDE PARAMETER AND TIME LIMITS AS APPROPRIATE]
Operating in accordance with FNP-1-UOP-1.:2 [Startup]
N41 indicates 2%
N42 indicates 4%
N43 indicates 5% i N44 indicates 3%
. Pressurizer pressure indicates 2135 psig }
j Pressurizer level indicates 48% '
Loop A Tavg indicates 540F '
Loop B Tavg indicates.542F j Loop C Tavg indicates 541F '
All SG indicates 33%
Rod Control in manual SG level control in manual ;
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8..-' ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE'60' ,
MAS!u.1l r !
QUESTION '8.07 (1.50)
Technical Specification 3.1.2.6, Borated Water Sources - Operating,
.' specifies that a boric acid storage system and refueling water storage 4 tank [RWST). borated water sources shall be OPERABLE.
A. What is-the Technical Specification boron concentrate'n o range for the RWST? [0.5]
B. If RWST solution temperature is less than the' minimum specified by j Technical Specification 3.1.2.6, what is the required Technical, '
specification action?
[ INCLUDE TIME AND LIMIT AS APPROPRIATE] [1.0]
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND; LIMITATIONS PAGE 61 l
WISlIR I i QUESTION:- 8.08 (1.50) w ws 3 A. ;How are. operations personnel notified of additions, deletions or irevisions to standing policies? , ;
q !
j B.. ;Where is the Operations Standing Policy Book maintained?
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- 8. ADMINISTRATIVE-PROCEDURES,'. CONDITIONS, AND' LIMITATIONS :i '
PAGE 62 MASTER
. QUESTION. 8;09 (0.75)-
~
s
-For an actual emergency it is deemed necessary to sound'the plant emergency
-- a l a rm .
.i What,is.the on?. minimum time that the emergency alarm should be allowed to stay '
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... 4 8 .'- ADMINISTRATIVE PROCEDUR'ES, CONDITIONS, AND LIMITATIONS PAGE 63 MASTER
-QUESTION. 8.10 (1,00) j When recorders?
should the chart' speed be noted on recorder charts for multi-speed
[ LIST TWO TIMES / CONDITIONS]-
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- 8. ADMINISTRATIVE PROCEDURES,. CONDITIONS,.AND LIMITATIONS- PAGE 64 1 1 MASTER QUESTICN 8.11 (l'. 5 0 )
-What'are the employee classifications designated as Tagging Officials?-
(LIST THREE CLASSIFICATIONS)
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8.. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE-65 i L
MASild
- QUESTION 8.'12 (1. 5 0 ),
What'are the Shift Supervisor's THREE. 3 position reported to and' reporting purp[os]e? basic reporting. chains including-
-[ LIST THREE (3) INCLUDING 1 POSITION REPORTED TO AND REPORTING PURPOSE]
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.8.. ADMINISTRATIVE ~PRGCEDURES,' CONDITIONS,'AND LIMITATIONS .PAGE 66.
I MASTER l
-QUESTION. 8.13 (3.00)-
- LIST SIX:[6] of-the responsibilities ~of the Shift. Supervisor as stated
'AP-16, CONDUCT.OF OPERATION - OPERATIONS GROUP, "
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS .PAGE 67 MASTER QUESTION 8.14 (2.50)
FARLEY - UNIT 1 Technical Specification 3.2.1, Axial Flux Difference is attached.
A. If the AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable, how often is AXIAL FLUX DIFFERENCE monitored and logged?
B. How long may operation continue at 45% thermal power with AXIAL FLUX DIFFERENCE indicating +5% before 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of penalty deviation accumulates? [ ASSUME INITIAL PENALTY DEVIATION IS 30 MINUTES]
C. How long may operation continue at 55% thermal power with AXIAL FLUX DIFFERENCE indicating -5%? [ ASSUME INITIAL PENALTY DEVIATION IS 30 MINUTES]
D. If AXIAL FLUX DIFFERENCE is -20% at 75% thermal power, what action should be taken to continue power operation? [ ASSUME INITIAL PENALTY DEVIATION IS 30 MINUTES] [1.0]
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- 8. ADMINISTRATIVE' PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 68 MASTER QUESTION 8.15 (2.00)
FARLEY - UNIT 1 Techni' cal Specification'3.3.3.9, FIRE, DETECTION
' INSTRUMENTATION, is attached .
A.
What action'should be taken if Room / Fire Zone 191, Auxiliary Fcedwater Pump Room, has no operable smoke detectors? (STATE ACTION AND LIMITS AS'APPROPIATE] '~
B. What action should be taken if Room / Fire Zone 55, Containment, has no operable Rate Test?smoke detectors while performing a Type A Containment Leakage I
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATI'ONS -PAGE 69 MASTER 1 i QUESTION '8.16 (3. 00)-
UNIT 1 IS AT 93%-LOAD The following equipment is out of service:
Charging' Pump A I Component Cooling Water Pump A l Residual Heat. Removal-Pump A Start up auxiliary transformer 1A Disel Generator 1-2A Component Cooling Water Pump B key interlock [ pump aligned to Bus 1F]
a An the Shift Supervisor, what action would you take?
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- 8. . ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 70 MASTER
~ QUESTION 8.17 (1.75)
FARLEY - UNIT 1 Technical Specification 3.7.1.5, MAIN STEAM LINE ISOLATION l VALVES, specifies that each main steam line isolation valve shall be OPERABLE in MODES 1, 2, and 3. What are the Technical Specification BASES for the main steam line isolation valve operability requirement?
)
(LIST / DISCUSS THE EVENT OF CONCERN AND TWO LIMITS / EFFECTS] j l
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS,'AND LIMITATIONS PAGE 71 MASTER QUESTION' 18 . 1 8 (1.00)
After transferring previously prepared (" batch'ed") boric acid from the BA Batch Tank (BABT) to storage tank "1B", the operator (according to procedure FNP-1-SOP-2.6) must:
- 1) . Verify 1A BAT Baric Acid Concentration is 4.0 to 4.4 w/o as indicated by latest chemistry' analysis.
- 2) Place 1A and 1B Boric Acid transfer pump control switches in-STOP and return tc AUTO
- 3) Shut batching tank supply to Boric Acid pumps 1-CVC-V-8310 (Q1E21V236).
Tha shift. supervisor has determined that it is desirable to perform steps TWO (2) and TliREE (3) prior to the " chemistry" verification called for in step ONE (1). You agree.
Assuming that the plant is in " normal" MODE ONE (1) operation and the
' system is functioning properly - what administrative guidelines would allow these procedural steps to be performed "out-of-sequence"? !
- STATE ANY ASSUMPTIONS TAKEN TO SUPPORT YOUR ANSWER -
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- . q 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS,'AND LIMITATIONS- PAGE 7:2 jj m QUESTION ,B.19 (1.00).
The only people allowed to manipulate any control that directly affects
.license reactivity AND or power ...: level are those who currently hold an NRC RO or SRO a) ..by those in training for an RO/SRO license IF; 1)' prior SRO approval has been received
- 2) individual is under direction of a licensed RO/SRO
- 3) individual is'in the presence of a licensed RO/SRO '
b) ..by those in training for an RO/SRO license IF;
- 1) prior SRO approval has been received
- 2) individual is under direction of a licensed RO/SRO !
c) ..by those in "non-licensed technical" positions IF;
- 1) prior SRO approval has been. received
- 2) individual is under direction of a licensed RO/SRO
- 3) individual is in the presence of a licensed RO/SRO
- 4) the job requires direct support from within the individual's field of expertise d) ..by those in "non-licensed technical" positions IF;
- 1) prior SRO approval has been received
- 2) tne job requires direct support from within the individual's i field of expertise l
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(************* END OF EXAMINATION *************)
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.I ATTACHMENTS FARLEY-1&2 SRO EXAM REGION II 88/11/14 I i
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MASTER ~
SYos + '
EQUATIOR SHEET.
O!1PY.
f = ma v = s/t Cycle efficieiicy = (Net work out)/(Energy in) 2 w = mg s = V,t'+ 1/2 at 2
E = mc . .
KE = 1/' c:v a = (Vf - V,)/t A = AN A = A,e'^
PE = agn Vf = V, + at w = e/t i = an2/t1/2 = 0.693/t1/2 t
,9 2 g,,f 1/2"##
- E(*U?)I"b)3 4 A= 4 ((t1/2)
- It b)3 AE = 931 om -
m = V,yAo ,
-Ex Q = mCpat 0u UAL T* I
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- Pwr = Wgah I = I,10"*/U L
. .. EL = 1.3/u .. ,
4 P = f q10 sur(t) HYL = -0.693/n t
P = PO e /T SUR = 26.'06/7 SCR = S/(1 - K,ff)
CR x = S/(1 X,ffx) .
50,c. = 26s/t* + (s - o)T CR;(1 - K ,ff)) = CR2 (1 - keff2) a T = ( t*/s ) + ({ s - o V Io ] M = 1/(1 -- K,ff) = CR)/CR, T = 1/(o - 8) , M = (1 - Keffo)/II ~ Keff1) l T = (8 - o)/(Io) SOM = ( - K ,ff)/K ,ff ;
o = (Keff-1)/K,ff = *,ff/K eff 1* = 10 seconds l
I = 0.1 seconds-l o = ((t*/(t K,ff)] + [a,ff (1 / + IT)] i Idjj=Id' P = (r4V)/(3 x 1010) Id j 2 =2Id 2 22 2
! t = cN R/hr~= (0.5 CE)/d (meters)
R/hr = 6 CE/d2 (f,,g)
Water Parameters Miscellaneous Conversions
]
1 gal. = 8.345 lbm. 1 curie = 3.7 x 1010ep, 1 gal. = 3.78 liters 1 kg a 2.21 lem 1 ft4 = 7.48 gal. 1 np = 2.54 x 103 Bru/nr Oensity = 62.4 lbm/ft3 1 mw = 3.41 x 100 Stu/hr j Oensity = 1 gm/cm3 '
lin = 2.54 cm l Heat of vaporization = 970 Stu/lem *F = 9/5'C + 32 Heat of fusion = 144 8tu/lbm 'C = 5/9 (*F-22) 1 Atm = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-lbf I 1 ft. H 2O = 0.4335 lbf/in.
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781 5 8MS 923 4 1001 0 1081 3 6158 9 1228 4 1291 4 13il l 1404 7 1453 9 IMUO IM38 1636 1 4 0 92M 0 9190 8 0JS8 109J9 I 1569 1 2060 4 2529 1 2946 I JJ71 1 3644 1 3931 4 4102 1 4671 i Sio0
$h = Superhe81. I h = enthalpy. Blu per Ib v = Specific volunit, tu fl per Ib $ = gr tropy, Blu per f per lb D -9
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MASTER COPY Table 3. Supe 7 healed Steam-Continued Aes Press . 19750 la Sal. ($al. femel - Sal - . I*"9"alisse-OessetsIthstaheel Walet $ lease - ilt lidet ..,
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- s. . Deres 5 -000767 9 0SLIS 0296 003M l599003868546 00:43 0043 00%? 0 0670 0 0676 19580 6
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)
t 0777 0743 0 0243 00?to 001'S 90370 0 7 sell 9424 984 % IM98 1114 9 (204 3 1268 7 1310 8 13844 14 08?3 0 0414 0 0534 0 0548 0 52000 e D 0 9163 0 M9f 1.0732 1 0777 3 4316 4 1840 l 2308 1.2727 1 3:07
,s 0 0248 00760 00?se 0 0317 00M7 004M 00e% 00508 00%0 O Mit SMS9 0 0704 776 8 Wld 9079 9778 IM09 lite t 1191 7 1258 0 1318 6 1374 7 4426 6 14751,1%
12300 e $9131 SMS7 ,10177 1 0708 11219 I lief i1209 l.2627 1.J010 8 JJ53 1 3467 1.394 h s ' -774 0 0?30 7 438 00?%90J 0 0779 090300 00344 - 00M0 0 0437 0IM)404s4talte 00SM l3005 h e 0 9101 0Mia6 10127 9 974 i M37 10438 lilii 1184
' I illi l 1453 I filil 12$34 I2479 IM64 1.2914 1 3264 :1 s 0771 0736 5 SM1 00PS390n4 0 077%
%64 10M $04P7 Sent 1106 ' 700176 1114 0 0470 50 17M 4 1734 044660t194 05129 .- 4l 13980 e 0 09073 ' 0M47 1 0000
' 0 0735 0 0751 0 0775 0 0797 00J10 00M4 10$18 11079 l 1%76 17030 t 744S $1831 4 s 777 3 834 4 8972 ' Mit 1030 0 1099 1 00405 00448 01.ast io6M ' 0M77 ' 80619 0 06 % 4 0764 1166 3 1779 7 1291 0 l344 6 1401 5 6451 4 Itati 18319 0 904) $ M48 1 0037 1 0524 ' 4 1014 114M 1.1946 3 2)61 1.2749 14 30 ,
6.
's 017t 0?33 3 832 0 0748 6 494 0 30?67 M80 0 0791 1074 5 01097 0370 3 00]S4 1134 501771 0192401701 04110 ON74 1340lJ lages n
0 9019 O MI) O h16 1 0473 1 9913 1 1426 1 1877 1 2781 12671 s 1 0170 0231 4 0 0746 431 0 3910 0764 7 M43 00787 1019 06 0114 1046 0014s 2 18514 0 0180 tills 0 041413379
.12744 0 0418130 13est ,
k- 0 8998 0 94W 0 9957 1 04?6 letti iIM2 i 1001 1!?0S I 2597 1 2949 s 00?)0 769 6 0- $195 0784 tail 0 0761 - 9% 00282 9 10li 00)of t 1080 ' 00317 6 1144 090369 1206 08040512681 0 0443 1316 00 isseg , 6 0 0970 0 94 % 0 9920 ' I 0382 1 0846 3 1302 1 1735 1 2139 I ful - s 0 0778 764 9 0 8'M6 R*a0? au; 0742 947300?$4 10:11 107scaf78 0 0'07 f 18390 170n001M3 126t l0lit 0340 96 13778 0 0393 14736 157040 0479 08 1410
' Sh - superheal. F 0 9477 096th 10J40 1 0797 6 1247 4 1674 5 7073 1 2457 178)$ 1 V = Sp0Clist volume, CU lt per lb h = enthalpy. Blu per Ib
$ = entJ0py, Blu per R pe_t Ib 9
8-10
l
.; / -
mP-1-ESP- 0.1\ Feactor Trip Response" MASTER Revision 6 Step Action / Expected Response Res T Obtained I I I I I
; 4 4 I l 4 4 4 4 CAttrION Ir establishing excess letdown, THEN l
' ma intain excess letdown heat exchanger outict temperature (TI-139) less than 1
********************************************65'r. ******
; I EXC LTDN ;
HX DISCH 't HIK-137 controller 1 throttled as required l l 7.3 Control charging flow tol -
' maintain pressurizer level l 19% - 24%.
PRza LVL -
- LK-459 controller adjusted !
as required CHG FLON ' I h FK-122 controller adjusted b - as required l 8 Check pressurizer pressure. 8.1 Check pressurizer pressure 8.1 Ir pressure Nttr purposely l greater than 1850 psig. Tiduced, THE U erify SI PI-455 actuation and go to
- mP-1-EEP-0, REAC'IDR TRIP PI-456 .
OR SAFETY INJECTION, Step 5. PI-457 SI AC'IUATION either switch to AC'IUATE if automatic actuation has laroccurred
\
~
I h t I t t t t t I l- 1 I l' i 15/32
\.
~
> i. b .
ENP-1-ESP-0.1 , Reactor Trip Respci e ' Revision 6
., m m -u n Step Action / Expected Response R e NOT Obtained
/ x 4 l 4 l 4 l 4 l l 4
~!4 CAUTION WHENEVER a pressurizer PORV opens ;
because of high pressure, THEN verification of proper PORV closure should be repeated after pressure falls to less than 2315
*****************************psig. l
********************* 1 8.2 WHEN pressure less than 8.2 Close associated block l 73T5 psig, THEN verify valve for any failed
, PORVs closed and no leakage pressurizer PORV.
evident. PRZR PWR OPER PRZR PONER OPER REL VLV RELIEF ISO VLV RC-PCV-444B closed RC-MOV-8000B closed if PCV-444B open RC-PCV-445A closed
~ RC-MOV-8000A closed if PCV-445A open POWER -
OPER (p 'T ret vov
- TI-463 < containment cooler ID intake temperature (TR-3188 (BOP]
point 8 or plant computer point hg TE31%7E) OR trending down PRT TI-471 < containment
~ lower compartncat temperature (TR-3188 [ BOP] -
point to or plant corputer point ica 43lBST.) OR trending down
~
l tA _ PI-472 < 10 psig OR stable OR trending ~ Town J t I 1 I t I t I t I t. I 16/32 RU M
f
*7 '
m71 h FNP-1-ESP-0.1 Reactor Trip Re lSeSTER Revision 6
'*2 -Step Action / Expected Response pp se NOT Obtained l l
'4 4 l l 4 4 l 4
8.3 Verify normal pressurizer 8.3 Ir normal pressurizer spray spray valves stabilizin l pressure at 2235 psig g Inoperable, TICN perform one of the following as .; PRZR PRESS appropriate. 4 REFERENCE I
~ PK-444A controller adjusted
- as required '
SPR VLV PVC-444C (D) 1
~ PK-444C controller adjusted as required
- PK-444D controller adjusted as required i
8.3.1 IF any normal pressurizer spray valve failed open, _THEN stop RCP.in i associated loops . g#- ,t ( ~ RCP 1A tripped for , PCV-444C failed open _RCP.lB tripped for i PCV-444D failed open
. 8.3.2 IF spray valves inoperable, THEN with letdown in l service use pressurizer auxiliary spray, otherwise !
use one pressurizer PORV i for reducing pressure when needed. - RCS ' PRZR AUX SPR CVC-HV-8145 used if letdown in service (FI-150 > 0 gpm) i PRZR PWR OPER i REL VLV l RC-PCV-4448 OR 445A l
-(but NOT bothJ used if {
auxiliary spray Unavailable l l g ii i i i 1 l l 1 i. 1 I I 17/32 '
- s. -
\ ,
.f .---
, i \
I
, mp-1-ESP-0.1 Reactor Trip Response Revision 6
). Step Action / Expected Response Itesponse NOT Obtained j l l l 4 4 4 l l
- 4 4 l l 8.4 Use pressurizer heaters 4 9
as needed for raising RCS pressure. fj PR4R HTR BKUP (VARIADLE) JU5 B _GRP 1A, B, C, C or E on or off as required
\
CAIRICN
\
IF CCNDENSATE S'IORAGE TANK LEVEL LO-LO ! MiAIN A (B) annunciator J-34 (J-44) alarms (setpoint 5.3 ft), THEN AEW pumps may be damaged if pump suction is not shifted to service water (refer to PNP-1-SOP-22.0 . 1
********************)****************************
9 Check steam generator levels.
'gg 9.1 SG narrow range levels i
W greater than 6%. 9.1 Maintain greater than 395 gpm total auxiliary
- LI-474, 475, 476 for SG 1A feedwater flow to SGs LI-484, 485, 486 for SG 1B until at least one SG
_~LI-494, 495, 496 for SG 1C narrow range level is greater than 6%.
-~ FI-3229A for SG 1A rI-3229B for SG 1B
[rI-3229CforSG1C 8 _SPDS display 2H1, HEAT SINK ! STATUS TREE if SPDS is available 9.2 Control AEW flow to maintain
~~
9.2 IF narrow range level in SG narrow range levels ~~ 6%-50%. any SG continues to rise, THEN stop feed to that SG.
~ LI-474, 475, 476 for SG 1A
~ LI-484, 485, 486 for SG 1B .
_LI-494, 495, 496 for SG 1C l l l 1
)
C4 t t t t t I I I t.
.I I i
18/32 l
. . . ~ . . _ _ . ~ . . . . . . . . .
1 s . . . .
+-
_FNP-1-FSP-0.1 Foldout Page Reactor Trio Restense Revisien 6 I ({ } Step Action / Expected Response I i gg T Obtained I f
- s. I no u t t t I Ti t ton 1 t i I Monitor RCP 'riteria, t f y
__ 1 1 SUB COOLED MARGIN MCNITOR 1.1 lbstonecharging w indicates greater than 16*F g (43'F}'subcooling in CETC pump running, THEN trip mode. all RCPs. 2 Monitor SI criteria. 2.1 Creater than 16'F (45'F) 2.1 Verify SI actuated. M subcooled in CETC mode
. " AND PRZR level above 7%
T5U%). [ w 3 Monitor CSF red path criteria. I 3.1 Nuclear power less than 5%. 3.1 Go to FNP-1-FRP-S.1. ! I 3.2 Core exit TCs less than 1200*F. 3.2 Go to FNP-1-FRP-C.1. 3.3 l
' Tocal A m flev greater than 3.3
'3S3 gpm OR at least one SG Go to mP-1-FRP-H.1. l
.h narrow range level' greater than 6% [34%).
3.4 RCS cold leg temperature greater 3.4 than 266*F OR cooldown rate less Go to mP-1-FRP-P.1.
- l than 100'F Tn last 60 minutes.
I 3.5 CTMT pressure less than 54 psig. 3.5 Go to FNF4-FRP-Z.1. l 4 _ Monitor switchover criteria. l 4.1 CST level above low low level 4.1 Shift A m pumps suction alarm (5.3 f't), to service water per FNP-1-SOP-22.C. I 5 Monitor charging miniflew criteria (6.'rina SI). l 5.1 RCS pressure less than 1900 psig. 5.1 verify miniflew valves open. l 5.2 RCS pressure greater than 1300 psig. 5.2 verify miniflow valves closed. 6 Monitor adverse containment criteria. 6.1 CTMT pressure < 4 psig ~AND 6.1 l 3 radiatien less than 10 R/hr.
~
Use adverse CET condition numbers ( }. l 7
) t t t Monitor FNP-0-EIP-9 TAB 3 criteria.
u i t i
- t i
t 1 32/32 yg I
~'-
- t. ,
(- FNP-0-EIP-12 July 3, 1985 Revision 5 MASTER FARLEY NUCLEAR PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE FNP-0-EIP-12 S A F E T Y ALERT C. RESTEICTED USE CCPY
' R E
b ' N' N L ISSUE DATE- A T EXPIRATION DATE: U IO-4P V/CC C2 3 I?O: gge gg,gg D
~ DC "-' ' ~2 I? CU"'?"NT nATE Approved: Ex a u.n....iic.s v a d
/,
l l- ['y j rt..;s Plant Nanager~ 1 Date Issued: List of Effective Pages l Page Rev. 1-7 5 Checklist 12A 5 Diskette #EIP-1 L
D 1 l
. . 4 1
t i FNP-0-EIP-12
.\ . ALERT ; 7" ' -(- 1. l .>
1.0 Purpose Ph
"] l This procedure defines the ' criteria fd/ yl s fying an i
emergency as an Alert, delineates personnel and organizations who may be notified and lists actions
.which may be taken to mitigate the effects of the emergency.
2.0 References 2.1 Joseph M. Farley Nuclear Plant Emergency Plan. !
- 2. 2 - FNP-0-EIP-8, Emergency Communications 2.3 FNP-0-EIP-9, Radiat' ion Exposure Estimation and
' Classification of Emergencies.
1 2.4 FNP-0-EIP-10, Evacuation and Personnel .
'l Accountability.
2.5. FNP-0-EIP-13, Fire Emergencies. 2.6 FNP-0-EIP-14, Re-entry Procedures. ( 2.7 FNP-0-EIP-26, Offsite Notification 3.0 General 3.1 Description The classification of Alert applies to situations in which events are in process or have occurred which involve an actual or potential substantial degradation of the. level of safety of the plant. The potential for release of-radioactive material for the Alert classification is up to 10 curies of I-131 equivalent or up to 104 curies of Xe-133 equivalent. The purpose of offsite alert is to assure that emergency personnel are readily available to respond if the situation becomes more serious or to perform confirmatory radiation monitoring if required and to provide offsite authorities current status information for possible further action. 3.2 Criteria An Alert would be declared for plant conditions that warrant precautionary activation of the ' technical support center, operations support
;( centers, and the emergency operations facility Rev. 5
. 1
- - _ - _ - - - _ _ _ _ - - . _ - - _ . - - i
b < I $ 1 l m; FNP-0-EIP-12
^ ,
B0PY (at the discretion of the Recovery Manager).
'( 'Specifically, an Alert would be declared for any of the following:
3.2.1 Severe less of fuel cladding as indicated. I by a reactor coolant activity equal to I or greater.than 300 pCi/ gram ee"' valent i I-131. ) ( 3.2.2 Steam generator tube rupture indicated l by: (a) ECCS actuation, AND (b) High secondary coolant activity (R-15, R-19, R-23A, or R-238 reach full scale). 3.2.3 Greater than 10 gpm primary to secondary leak as indicated by high secondary. coolant act.vity (R-15, R-19, R-23A, or R-23B alarving) OR main steam isolation valve failure with a steam line break outside containment indicated by: (a) Abnormally low steam pressure on !
. /- one or all steam generators with .(.
one or more of the following: (1) Steam line high flow OR i (2) Steam line high differential pressure OR (3) Steam flow greater than feedwater flow i i AND (b) No abnormal temperature, or humidity { < increase in containment. 3.2.4 A primary coolant leak greater than 50 gpm. Indications of such a leak will include high charging flow AND (a) High containment radiation (R-2, R-22, and R-12) AND (b) High containment humidity. l OR_ . (c) Pressurizer relief or safety valve discharge line temperature high ANE
-{ (d) Pressurizer relief tank level, pressure or temperature increasing or above normal. 1 Rev. 5 {
2 j
I % FNP-0-EIP-12 3.2.5 High radiation levels or high airbot:ne ( contamination indicative of a severe degradation in the control of radio-active materials as indicated by:
) (a) Readings on R-14 (stack gas monitor),
' ,i R-21 (stack particulate monitor) OR R-22 (stack gas monitor) reading pg y off scale, bW l Ano ,
(b) Sampling or R-27 high range offluent l monitor confirms direct readings.
- 3. 2.6 Loss of offsite power with a failure of all emergency AC power for less than 15 minutes.
3.2.7 Loss of both trains of auxiliary building DC power for less than 15 minutes. 3.2.8 Loss of both trains of: (a) Auxiliary feedwater (Modes 1-3), OR (b) RHR (All modes), OR l (c) CCW (modes 1-4), OR I (d) Service Water (modes 1-4) I 3.2.9 Spent fuel handling accident in which an increase in radiation level (i.e., alarm - condition or off-scale reading) is observed ott R-2, R-11, R-12, R-5, OR R-25 as a result of one of the following: (a) Dropped spent fuel assembly, OR (b) An object is dropped onto a spent fuel assembly, OR (c) A cask containing a spent fuel assembly is dropped, OR (d) A spent fuel assembly is deformed as a result of any manipulation, OR (e) Low spent fuel pool water level. 3.2.10 Loss of all main control board annunciator capability. Rev. 5 3
\
o p , g <
,, li: y s-
{ .' 1 , i J{d 3.2.11' Radiological effluent at the site boundary
- s. 3
-j (combined effect from bothrunits): greater '
( l than 10 times the radiological technical
'[ specification ~ instantaneous limits as
+-
, l follows:
J 3.2.11.1 Liquids: 10 times 10CFR20
; Appendix B Table II Column 2-3 ? _. 2 Dissolved or' entrained noble gasses: 0.002 pci/ml 3.2.11.3 Noble gasses (whole body):
0257 mrem /hr i ! 3.2.11.4 Noble gasses (skin): 3.4 mrem /hr 3.2.11.5 Airborne radioiodine'and , s particulate other than noble gasses: 1.7 mrem /hr NOTE: The limit in 3.2.11.5 is less conservative (. than State of Alabama g limit in 3.2.12 below. 3.2.12 Projected exposure-rats at site boundary for manual calculations or projected peak dose rate location within the gaseous. effluent plume for automated emergency dose calculation method (EDCM) calculation: A. 2 1 mr/hr whole body' exposure CA B. t 1 mr/hr thyroid exposure 3.2.13 A security emergency involving the ! I occurence of or imminent threat of sabotage. 3.2.14 Severe natural phenomena being experienced l or projected as follows: (a) Ea'rthquake greater than \SSE levels (0.05g ground acceleration) k, Rev. S 4
+ x
; y: y .. ~i '
1 i
]bl, kI 4 IhM
.FNP-0-E2P-12 10PY 1 1
' :/
-(? ,
i(b)' Flood, low river water or hurricane-surge near; design levels that could , ,
. impact plant operations.
3 J 1 l
- 1. (c)- Any tornado. striking; facility '
[ , 3 (d). Hurricane' winds near design' basis 1 j level (115 mph)- - 3.2.15 ' Fire or. explosion potentially affecting l
! ,.ECCS
\'
3.2.16l' Failure of the reactor protection system 'l l to initiate and complete a trip which' ; i brings the reactor subcritical. 3.2.17 A acsteam or feed indicated line break'inside-containment by abnormally low' pressure l!
, on one steam generator with the following: j i
(a) Steam line high differential pressure,'
- 98. l l (b) Staam flow greater than feed flow, EB !
(c) (' Steam line high flow, AND (d) Containment high temperature. . 3.2.18 ' Single rod cluster control assembly I withdrawal'at' power as detected by: (a) Rod position indicator, AND (b) Increasing core power, AND (c) Increasing Tavg. 3.2.19 Hazards experienced onsite which affect 1 4 plant operation cuch as-(a) Aircraft crash
\
(b) ,Release of toxic gas
- ' -~
(c) Release of flammable gas ~' 3.2.20 Coolant pump seizure leading to fuel failure as indicated by radiation monitors or gross failed fuel detectors. m Rev. 5 5
'\
- NASTER s .
1 FNP-0-EIP-12 9 pg J L&g}i E l 3.2.21 Evacuation of control room anticipated ! (- or required with control of shutdown systems established from local stations. l 4.0 Procedure i 4.1 The Shift Supervisor will. perform the following: 4.1.1 Announce the condition over the plant public address system if appropriate. 4.1.2 Evacuate affected areas of the plant as appropriate. l 4.1.3 Implement notifications per EIP-26 I 4.1.4 The Shift Supervisor shall perform the . I 5 duties of the Emergency Director until his arrival and assumption of duties. 4.2 The Emergency Director shall perform the folloving: 4.2.1 Activate appropriate portions or plant l Emergency Organization. 4.2.2 Announce the condition over the plant ! ( 4.2.3 public address system if appropriate. Evacuate affected areas of the plant as appropriate. I l 4.2.4 Notify the proper offsite authorities : per EIP-26.
.r 4.2.5 Activate the Technical Support Center I and Operational Support Centers and EOF (at the discretion of the Recovery Manager) to the extent required to respond to conditions precipitating the Alert.
4.2.6 Dispatbh a Radiation Monitoring Team if I a release is imminent or in progress. 4.2.7 Provide periodic meteorological and dose i estimates and release projections based on plant conditions and foreseeable contingencies to offsite authorities. 4.2.8 Plan and initiate re-entry (EIP-14) i 4.2.9 Provide periodic plant status updates to I offsite authorities. Rev. 5 6
l MASTER FNP-0-EIP-12 4.2.10 Continually reassess the emergency I (' condition to ensure that a higher j classification does not exist. t 4.2.11 If a fire, implement EIP-13, Fire l Emergencies. 4.2.12 Close out by verbal summary to offsite l l l, authorities followed by a report as required by technical specifications or ) escalate to a higher level emergency. j D L Rev. 5 l 7
1 i-
'. Pb s.8 e FNP-0-EIP-12A i
.i l
J( ALERT Initials I. Shift Supervisor A. Evacuate affected areas of the ! site as necessary i B. . Implement EIP-26 II. Emergency Director A. Activate appropriate! portions of the plant emergency organization. B. Evacuate area of plant site, as l'
-necessary I C. Notify offsite authorities per EIP-26 l; D. Activate-TSC, OSC, and EOF (at- <
l~ 1 Recovery Manager's discretion) ~ j E. Initiate environmental sampling
; q.
.l !
.( F. Provide periodic meteorological and dose estimates to offsite authorities l G. Plan and initiate re-entries 1 (EIP-14)
H. Provide periodic plant status updates to offsite authorities -l I. Reassess conditions for possible upgrading of. emergency classification J. If a fire, implement EIP-13 K. Close out by verbal summary or' escalate to a higher level emergency
=
( . 1 Rev. 5
1 . o,- i ;; { 1, 1 l l M) FNP-0-EIP-17 8 ['... i ' $$ l6 July 3, 1985 ' Revision 4 00PY FARLEY NUCLEAR PLANT 4 1 l EMERGENCY PLAN IMPLEMENTING PROCFDURE FNP-0-EIP-17 I s i A : F. i
, E T
Y NOTIFICATION OF UNUSUAL EVENT _ .- ...- my ;.2 CC W '
- _R
. , ; :_ & .ll-M E
g
. li-3 N --
. . A T
. . ._ _ .: 0:M
. E D-
;:5 .
"~~'~~"~
Approved: '._.
' '\
- i. ** . rY Plant Man &ger Date Issued: ? ,, ' . i ; '
Diskette #EIP-1 List of Effective Pages Page Rev. ."' N 1-4 4 Checklist 17A 4-
. .? . . . .. . . . -
FNP-0-EIP-17 ! A NOTIFICATION OF UNUSUAL EVENT
- 1.0 P,u, rcose This procedure defines the criteria for classifying an emergency as a Notification of Unusual Event, delineates personnel and organizations who may be notified and lists actions' which may be taken to mitigate the effects of the emergency.
2.0 References i 2.1 Joseph M. Farley Nuclear Plant Emergency Pl$.n. 2.2 FNP-0-EIP-8, Emergency Commu!aications ; l 2.3 FNP-0-EIP-13, Fire Emergencies i 2.4- FNP-0-EIP-11', Handling of Injured Personnel 2.5 FNP-0-CIP-26, Offsite Notification 3.0 Ceneral 3.1 Description The classification of Notification of Unusual ; Event applies to situations in which events are ' in process or have occurred which could indicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety
. systems occur.
I 3.2 Criteria I A NOTIFICATION OF UNUSUAL EVENT would be required for any plant condition that warrants increased awareness on the part of state and/or local l offsite authorities or involve other-than-normal j plant shutdown or any of the following items: 3.2.1 Initiation of safety injection either ! automatically or manually as a result of plant parameters apli roaching or reaching their satpoint. 3.2.2 Radiological effluent at the site boundary } (combined effect from both units) in excess of the radiological technical specification instantaneous limits as follows: 3.2.2.1 Liquids: 10CFR20 Appendix B, ' Table II Column 2 Rev. 4 1
.m .
. . . . .a. . . . . .
_.a. ' L , . , . .'-. N %,* L FNP-0-EIP-17 ' 30PY 3.2.2.2 Dissolved or. entrained noble gases: 0.0002 pci/ml l 3 2.2.3' Noble gasses (whole. body): O.057 mrem /hr 3.2.2.4 Noble gasses (skin): 0.34 mrem /hr 3.2.2.5 Airborne radiciodine and particulate other than noble l gasses: 0.17 mrem /hr-L 3.2.3 Any of the following technical specification limiting conditions for operation are exceeded: (a) Reactor coolant activity requiring
-l .
shutdown. ! (b) Loss of containment integrity ! requiring shutdown to HOT SHUTDOWN. 3.2.4 Indicated'subcooling (margin to-saturation) decreased below 100F. i , 3.2. 5 Fail'ure 'of any of. the following valves l [ i s L
.to close:
i f (a); Pressurizer safety valve, l l (b)! Pressurizer.poweroperatedrelief I valve and'its remote motor operated Iisolation valve. 4 (c) A steam generator safety valve, (d) A steam generator power operated relief valve ~. 3.2.6 Loss of both trains of offsite power OR i loss of all onsite emergency power - (diesel generators and auxiliaries). 1 3.2.7 Attempted unauthorized entry into a vital area or attempted sabotage of vital equipment. l 3.2.8 Loss of control room indication or annunciation to an extent requiring shutdown. 3.2.9 Natural phenomena being experienced or projected to affect the plant site as follows: 2 Rev. 4
y , n l MASTER FNP-0-EIP ! i 30PY (a) Any earthquake, (b) Unusual river water level caused by flood, low water or hurricane-surge, (c) Any tornado onsite, (d) Any threatening hurricane. ! 3.2.10 Hazards experienced onsite or within one mile of the site boundary which could affect plant operations, such as: (a) Aircraft crash, (b) Explosion or fire, (c) Release of toxic gas, (d) Release of flammable gas. I 3 2.11 Transportation of contaminated injured !
-individual to an offsite facility. !
, 3.2.12 Loss of secondary coolant outside containment
-concurrent:With ECCS activation.
3.2.13 Complete loss of forced .RCS flow as indicated by RCS flow indicators on all s three RCS loops. 3.2.14 Inadvertent loading of a fuel assembly into an improper position which causes Fq to be greater than the technical I specification limit. 4.0 Procedure 4.1 A plant operator shall notify the Emergency Director of any emergency if the Shift Supervisor is indisposed. 4.2 The Shift Supervisor will perform the following: 4.2.1 Evacuate affected areas of the plant as applicable. 4.2.2 Notify the individuals or authorities as delineated in EIP-26. Rev. 4. 3
R
.e p
FNP-0-E'IP.-17 4.2.3 The shift Supervisor shall perform the duties of the Emergency Director luntil his arrival or as otherwise instructed, by the Emergency Director. 4.3 The Emergency Director will perform the followingi I 4.3.1 Notify the offsite: authorities i delineated in EIP-26. I~ l ! 4.3.2 Activate appropriate' portions of the o plant ~ emergency organization. 4.3.3- Ensure personnel' accountability if an evacuation was initiated (EIP-10). l 4.3.4 Plan and initiate:re-entries (EIP-14). 4.3.5 continually reassess the emergency condition to ensure a higher classification does not exist. 4.3.6 Close out with verbal notification to notified agencies followed by a written report as required by technical specifications or escalate to a more severe class. i!
}
l 4 Rev. 4 1
r ' g -, e o h* [ d
. Fg gg.1[ FNP-0-EIP-17A L.# WI. I NOTIFICATION OF UNUSUAL EVENT CHECKLIST Initials I. Shift Supervisor A. Announce condition and initiate evacuation if required.
, B. Notify individuals / agencies per *
. EIP-26 II. Emergency Director A. Notify offsite authorities per EIP-26 B. Activate appropriate portions of the plant emergency organization.
C. Ensure personnel accountability j (EIP-10) D. Plan and initiate re-entry (EIP-14) l ! E. Reassess conditions for possible l l upgrade to a higher emergency classification l F. Close out or escalate to a more l l severe class l
s i u . c . l ENP-0-EIP-18 - l May 7, 1987 Revision 7 b
- y ,-
i , bh .im s m30EY FARLEY NUCIIAR PLANT ' 1 EMERGENCY PLAN IMPLEMENTING PROCEDUPI l . l. i FNP-0-EIP-18 - J i l 5 1 A 1 1 T E T Y SITE AREA EMERGENCY ! R E REST 2ICTED USE COPY L A ISSLT DATE- S - Il - MP -. T E E::.::nT:c:t carg._H-30 -ff ._ o 4 proved. L C:C:C~1IiCsV4C B44 i
/ Ed "OT U{I '" C"L.F.T T DATE 1
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Gene ~ral Manager - Nuclear Plant
%/ Date Issued:
[, , y - pg
/
List of Effective Pages Pace Rev. I T~ T Checklist 18A 7
'I EIP-1B/5 1
O 4 e 4
V '
, ,,. V 9 l b " U l7
- FNP-0-EIP-18
!L M SITE AREA EMERGDMUE _ E 1.0 Purpose I L This procedure defines the criteria for classifying an emergency as a l Site Area Emergency, delineates personnel and organizations who may. be notified and lists actions which may be taken to mitigate the l effects of the emergency. 1 ! 2.0 References l' . 2.1 Joseph M. Farley Nuclear. Plant Emergency Plan. 2.2 INP-0-EIP-8, Notification Roster. l 2.3 FNP-0-EIP-9, Radiation Exposure Estimation and Classification of Emergencies. 2.4 FNP-0-EIP-10, Evacuation and Personnel Accountability. 2.5 INP-0-EIP-13, Fire Emergencies. ! 2.6 INP-0-EIP-14, Re-entry Procedures. 1 2.7 INP-0-EIP-26, Offsite Notification i 3.0 General 3.1 Description ne classification of Site Area Emergency applies to those events which are in progress or have occurred that involve actual or likely major failures of plant functions needed for protection of the public from radiation or contamination. We potential for release of radioactive material for the Site Area Emergyncy classification is up to 1000 Ci of I-131 equivalent or 10 to 10' Ci of Xe-133 equivalent. 'The purpose of the , declaration of a Site Area Emergency is to: ! (a) Assure that response centers are manned, (b) Assure that monitoring teams are dispatched, (c) Assure that personnel involved in an evacuation effort of near-site areas are at their duty stations if the . situation worsens, and, 1 (d) Provide current information for and consultation with ' offsite authorities and the public. 3.2 Criteria A Site Area Emergency would be declared for plant conditions ( that warrant activation of emergency centers and monitoring ! 1 Gen. Rev. 7 1 _ __ __ 1
~
3
. MASTER FNP-0-EIP-18 .
teams. Specifically a' Site Area Emergency would be declared' for any of the following: 3.2.1 A major loss of primary coolant as indicated.by: (a) ' Decreasing pressurizer pressure and possible i
. level, AND ,
(b) Near normal steam pressure in all steam generators accompanied by,.
.(1) Containment pressure reaching 27 psig, AND (2) ' High containment radiation (R-2, R-il, and ,
R-12 reaching their alarm setpoint), AND ! (3) High containment sump (recirculation) leveli , i _AND (4)' High containment humidity. 3.2.2 Degraded core conditions with possible loss of core .! geometry as indicated by: (a) AT between RCS wide ~ range hot' leg and cold leg temperature >64*r and core' exit temperature (in-core thermocouple) reading greater.than 800*r and increasing, _0R (b) Core exit temperature (in-core thermocouple).
>l200*F.
3.2.3 A loss of offsite power and a steam generator tube rupture as indicated trf: (a) ECCS actuation, g (b) High secondary coolant activity (R-15 or R-19
' reach full scale) 3.2.4 Greater than 50 gpm primary to secondary leak, fuel' damage as evidenced by a reactor coolant activity greater than technical specifications, and a steam line break outside containment as indicated by:
) (a) Abnormally low steam pressure on one or all steam generators with one or more of the following: (1) steam line high flow,- , _l (2) Steam line high differential pressure, ' (3) steam flow greater than feed flow 2 Gen. Rev. 7
\
I
r -- , FNP-0-EIP-18 i _ 40PY
-(b) . _No abnormal. temperature or humidity increase in containment, i,
- . 3.2.5. Loss of offsite power.with a failure of all emergency L
AC power for more than~15 minutes. - 3.2.6 Loss of both trains of auxiliary building DC power for a more than 15 minutes. 1 3.2.7 Loss of functions for achieving hot standby. ; i 3.2.8 Spent fuel handling accident for which sampling or radiation monitors indicate a projected lower limit of
.offsite individual exposure to be:
1.0 Rem - Whole Body or i 2.5 Rem Thyroid .l 1 as a result of onen of the following:
-l (a) Dropped spent fuel assembly, OR (b). An object is dropped onto a spent fuel' assembly,.
OR t
-(c) A cask containing a spent fuel assembly is l dropped exposing the assembly, OR J (d) A spent fuel assembly'is deformed as a. result of any manipulation, OR (e) Spent fuel pool water level below top of I assemblies.
3.2.9 A' fire affecting ECCS. 3.2.10 Loss of all n:ain control board annunciator capability for more tEan 15 minutes while:: (a) Plant is not in cold shutdown, OR (b) Significant plant transient is initiated while ' all alarms lost. 3.2.11 Imminent loss of physical control of the plant (i.e., l takeurer by terrorists, anti-nuclear factions, etc.). f i 3.2.12 Severe natural phenomena being experienced or projected with plant not in cold shutdown: 1 (a) Earthquake greater than SSE levels 3 Gen. Rev. 7 ,\.
l.: . ,j l ' ,. .' i FNP-0-EIP-18 . J COPY
. Flood, low river water, or. hurricane surge -,
1 (b) greater than design levels. I (c). Winds in excess of 113 mph.-
'3.2.13 Other hazards being exper!.enced with the plant not in cold shutdown ~as follows.
.(a) Aircraft crash affecting vital structures by fire or impact, OR,
. 'i
-(b) Severe damage.to safe shutdown equipnent from !
missiles or explosion, OR (c) Entry of toxic or flanmable gases into' vital-areas. a 3.2.14 Evacuation of the control' room and control of. shutdown. systems not established <from local stations in 15- j minutes. 3.2.15 Any event for which sampling or radiation monitors indicate a projected offsite individual exposure to be greater than or equal to:
-1.0 Rem - Whole Body '
q E , j 2.5 Rem - Thyroid 3.2.16 Rupture.of a control rod mechanism housing as indicated by the following:
-(a) Rod position indication,.AND ,
-(b) High RCS pressure surge, g (c) Momentary nuclear power surge, g (d) Subsequent behavior indicating a loss of primary ,
coolant, _ l l 3.2.17 Transients requiring operation of sh'utdown systems with failure to trip (continued power generation but no core ~ 1 damage immediately evident). l ; 4.0 Procedure 4.1 The Shift Supervisor shall perform the following: l 4 . Rev. 7 - _: = .-
A" '
.FNP-0-EIP-18
'4.1.1 _ Sound the Plant Emergency Alarm and announce the condition and give needed evacuation. instructions over the plant public address system.
4.1.2: . Implement EIP-26, Offsite Notification 4.1.3 If emergency is a fire, also refer to EIP-13, Fire
-Emergency ,
4.1.4- 'Ihe Shift Supervisor shall perform the duties of the Emergency Director until his arrival and assumption of duties. 4.2 The Emergency Director shall' perform the following:
, 4.2.1 Activate appropriate portions of the plant emergency organization to include the Technical. Support Center, Operations Support Centers, and Emergency Operations . 'l Facility,'as necessary. '
4.2.2 Implement notifications per EIP 26. 4.2.3 Ensure personnel accountability (EIP-10). 4.2.4 Plan and initiate re-entries per EIP-14. 4.2.5 Dispatch Radiation Monitoring Teams. If additional support is required refer to EIP-8.
'4.2.6 Provide periodic meteorological and dose estimates and 1 release projections based on plant conditions and j foreseeable contingencies to offsite authorities. !
4.2.7 Provide periodic plant status updates to offsite l authorities. o i 4.2.8 Coordinate with the Recovery Manager with respect to information to be released to the press and recovery j planning. 3 4.2.9 Coordinate with the Recovery Manager to rend a corpany representative to the Houston County Centra.'. Emergency. Operations Center (CEOC). i 4.2.10 Continually reassess the emergency and change emergency classification as needed. l 4.2.11 Close out or recommend reduction in emergency class by briefing of offsite authorities and by phone followed ) s by written report as required by technical
, specifications; or escalate to a General Emergency.
5 Gen. Rev. 7
I a . g(* gy ENP-0-EIP-16A bli I - SITE AREA EMERGENCY CHECKLIST - Initials
-Shift Supervisor
-I.
. A .' Sound PEA, if necessary announce condition and give evacuation instructions B. Implement EIP-26 7I .' Emergency Director A. Activate appropriate portions of the plant emergency organization.
B. Implement EIP-26 l C. Ensure personnel accountability (EIP-10). D. Plan and initiate reentries (EIP-14). E. Initiate environmental sampling via RMT's. . i Provide periodic meteorological and F. dose estimates and release projections based on
. plant conditions and foreseeable contingencies to offsite authorities.
G. Provide periodic plant status updates to offsite authorities. H. Coordinate with Recovery Manager with respect to information to be released to the press and recovery planning. I. Coordinate with Recovery Manager sending company representative to . Houston County CEOC. ! J. Reassess conditions for possible change of emergency classification . K. Close out emergency with proper authorities. i Checklist 18A l EIP-1B/6 Rev. 7
?! FNP-0-EIP-19 Q, January 23, 1987 q Revision 8 l l
l 1 FARLEY NUCLEAR PLANT I EMERGENCY PLAN IMPLEMENTION PROCEDURE f FNR-0-EIP-19 l S A F E T i Y GENERAL EMERGENCY
'I R
E RESTRICTED USE COM " (SCUE DATE:- c-It-JF _- A T E Cr: :RATICU DATE: u-30 -id D ! W;E ORDER NO: U - Approved: i
-
- i r.- M r ,' . / i,1_cT;j t* 2 !? C" _:.ENT
._u!RATIC.. EATE DATE
! j .> / .,
GenerRManager - Nuclear Plant v Date Issued: ( f
,e. -
f
. i l
List of Effective Pagas ! Page Rev. 1-5 8 Checklist 19A 7 Figure 7 EIP-4/14 i______.._._ _ ._ _ .
,1
.FNP-0-EIP-19 j
GENERAL EMERGENCY l 1.0 Purpose-This procedures ~ defines the criteria for classifying an emergency as a General Emergency, delineates personnel and organizations who may be notified and-lists actions which may be taken to mitigate the effects of the emergency. , 2.0 References 1 2.1 Joseph M. Farley Nuclear Plant Emergency Plan, 2.2 INP-0-EIP-8, Emergency Communications. 2.3 INP-0-EIP-9, Radiation Exposure Estimation and Classification of emergencies.
- i i
2.4 INP-0-EIP-10, Evacuation and Personnel Accountability. 2.5 INP-0-EIP-14, Re-entry Procedures. I i 2.6 INP-0-EIP-26, Offsite Notification 3.0 General
'3.1 Description The classification of General Emergency applies to those events which are in progress or have occurred which involve actual or j iminent substantial core degradation or melting with potential loss of containment integrity. The potential for release of radioactive material for the General Emergency classification ~,
is more than 1000 Ci of I-131' equivalent or more than 10*. Ci of-Xe-133 equivalent. The purpose of the declaration of a General Emergency is to: (a) Initiace predetermined protective actions for the public. (b) Provide continuous assessment of informat! ion from licensee and offsite measurement. (c) Initiate additional measures as indicated by event releases or potential releases and, 1 Gen. Rev. 8 l l _ _ - _ _ - _ _ - _ - _. -. t
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-(d) Provide current information for and consultation with offsite authorities and the public.
3.2- Criteria-3.2.1 sampling indicates a projected offsite individual exposure to be greater than or equal to: 5 Rem - Whole Body ~O_R , b 10 Rem - Thyroid under actual meteorological conditions. 3.2.2 Loss of two of three fission product barriers with a potential loss of the third. The following describe
-indication of loss of these boundaries:
3.2.2.1 ruel cladding damage indicated by RCS-activity or loss of core geometry is indicated by AT between RCS wide range hot leg and cold leg temperature of >64* r and core exit temperature (incere thermocouple). y reading greater than 1200* r. ' 3.2.2.2 . Loss of primary coolant boundary as indicated by: (a) Containment pressure reaching 27 psig AND 1 (b) High containment radiation (R-2, R-22 ! and R-12, reaching their alarm setpoint) 1 _AN_ ,D. ; (c) High containment humidity. 3.2.2.3 Loss or potential loss of containment integrity is indicated by: ; (a) Containment pressure greater than 54 psig, OR- , 1 (b) A rapid decrease in containment pressure, OR l (c) railure of the containment isolation system resulting in a direct path from containment to the environment. 2 Gen. Rev. 8
. . -. c .. . . . .
FNP-0-EIP-19 4 3.2.3 Loss of physical control of the facility. 3.2.4 Other plant conditions exist, from whatever source, ' that make release of large amounts of radioactivity in '. a short. time period possible, such as any core melt !
., situation.
' 4.0 Procedure )
j
,4 .1 The . Shift Supervisor shall:
4.1.1 Sound the Plant Emergency Alarm and announce the corxiition and give evacuation instructions, if necessary, over the plant public address sytem. 4.1.2 Implement notifications per EIP-26. i 4.1.3 'Ihe Shif t Supervisor shall perform the duties of the Emergency Director until his arrival and assumption of duties. 4.2 The Eme':gency Director shall perfonn the following: 4.2.1 Notify the following state and local agencies via the DN. If at least one agency in each state does not I acknowledge within 10 .ainutes notify at least one agency in each state per EIP-8. 4.2.1.1 Alabama Bureau of Radiological Health (non-continuous staffing) or Alabama Department of Public Safety (continuous staffing) 4.2.1.2 Houston County Alabama Dnergency Management Agency (non-continuous staffing) or Houston County Alabama Sheriff Dispatcher (ccntinuous staffing) . 4.2.1.3 Georgia Department of Natural Resources (non-continuous staffing) or Georgia Emergency Management Agency.(continuous staffing) 4.2.1.4 Early County Georgia Civil Defense (aon-continuous staffing) or Early Ccunty Georgia Sheriff Dispatcher (continuous ' staffir.9) 3 Gen. Rev. 8
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o&'- F MASTER . hhf h: , .fr q g k/g FNP-0-EIP-19? < -. i
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o i ,, '4.2.1.5 Recommend to the agencies innediate i
* $' evacuation or shelter for all of'the general ,
population within,a two alle radius of FNP- 1 , (Zone A) and five miles down wind of FUP, unless more extensive protective actions are
'. i < ,
known to be. required.
~
- a. Specify affected zonss~when making J I j j
recommendations (See rigure).
, j 9
x . b.L When specifying evacuation zones, h consideration should be given to sectors L adjacent to the plume location.
! i
,, c. Wind variability should be considered 3[, when selecting the width of evacuation. '
T zones.. h' . . d. Disregard those portions of 10 mile
,L '
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zones which fall within 3 miles of l'
.INP when evacuating to G mile radius'.
4.2.2' Implement notifications per EIP-26.
'4.2.3' Activate appropriate portions of.the plant emergency. '
organization to include the Technical Support Center. 1 (TSC), Operations Supy rt Centers (OSC), and Emergency- i operations racility (Eor).. i 4.2.4 Ensure personnel accountability (EIP-10). j i 4.2.5 Plan and initiate re-entries per EIP-14. I i 4.2.6 Dispatch Radiation Monitoring Teams. If additional !
/ Support is: required, refer to EIP-8. i 4.2.7 Provide periodic meteorological and dose estimates and. ]
release projections based on plant conditions and foreseeable contingencies to offsite a'uthorities. 4.2.8 Provide informatic'n to the Recovery Manager with-respect to information to be released to the press and i recovery planning. 4.2.9
]
In conjunction with the Recovery Manager provide fer dispatching personnel to principal government agencies, I as necessary. 3 4 Rev. 8 u
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i ' i,,. f. , i FNP-0-EIP-19 4.2.10 ' Provide periodic plant status updates.to offsite l authorities. , i
- 4.2.11 Close out or recommend reduction ofl emergency class by briefing of off-site authorities by phone followed by- '
written report as required by technical specifications. j - fd
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-GENERAL EMERGENCY CHECKLIST i i
I. Shift Supervisors . , , A.- Sound PEA, announce'condit d give evacuation instruction,gif j necessary.
;y B. Notify Emergency Director _
C. Notify Security if necessary. II. Emergency Director A. Notify state-and local agencies
~
B.- Implement EIP-26 p C.- Activate appropriate-portions of'the plant emergency organization,to'
~ include TSC, OSC's, EOF. '
D. Ensure'personnal accountability '
-(EIP-10).
E.- Plan and initiate re-entries - (EIP-14). F. Initiate environmental sampling a via RMT's ' I i G. Provide meteorological and doce . estimates to off-site authorities.-
~
.l H. Coordinate with Recovery Manage'r on .
l press releases and recovery planning. ! I. Provide for dispatching of company representative to off-site agencies. J. Provide periodic plant status updated I to offsite authorities j K. Close out or reduce classification i of emergency. l
.g
,1 Rev. 7 r- ,
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. . 1 *.
3/4.2 POWER _ DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX 1 MA^ STi EiR "FERENCE (AFD) l' LIMITING CONDITION FOR OPERATION' i i L3.2.1 '1he indicated AXIAL FLUX DIFFERENCE ned within a rence. (A i APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER
- ACTION: I a.
With the indicated AXIAL FLUX DIFFERENCE outside of the band about the target flux difference and with THERMAL POWER:5% ta 1. Above 90% of RATED THERMAL POWER, within 15 minutes: a) Either limits, restore or the indicated AFD to within the target band ' b) Reduce THERMAL POWER to less than 90% of RATE POWER. i 4 2. I Between 50% and 90% of RATED THERMAL POWER: a) POWER OPERATION may continue provided: 1) The indicated AFD has not been outside of5%the target band for more than 1 hour penalty deviation cumulative during the previous 24 hours, and
- 2) i The indicated Figure 3.2-1. AFD is within the limits shown on Otherwise, reduce THERMAL POWER to less than 50% of RATED THERMAL' POWER within and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED i THERMAL POWER within the next 4 hours.
b) Surveillance testing of the Powdr Range Neutron Flux Channels may be performed pursuant to Specification 4.3.1.1 provided the indicated AFD is maintained , within the limits of Figure 3.2-1. ! A total of 16 hours operation may be accumulated with the AFD outside of the
- b. target band during this testing without cenalty deviation, POWER unless the indicated ACTION a.2.a) 1), above has been satisfied.
AFD 5% target is within the band and "See Special Test Exception 3.10.2 l
\.
t FARLEY-UNIT 1 ! 3/4 2-1 AMEN 0 MENT NO. 26 m
t POWER DISTRIBUTION LIMITS ACTION (Continued) c. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD has not been outside of the 15% target band previousfor 24 more than -1 hour penalty deviation cumulative during the hours. Power increases beyond 50% of RATED THERMAL POWER du not require being within the target band provided the
. accumulated penalty deviation is not violated.
SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within-its limits during POWER OPERATION 'above 15% of RATED THEREL POWER by: a. Monitoring the indicated AFD for each OPERABLE excore channel: 1. At and least once per 7 days when the AFD Monitor Alarm is OPERABLE, 2. At least once per hour for the first 24 hours after restoring the AFD Monitor Alarm to OPERA 8LE status. ' b. Monitoring and logging the indicated AXIAL FLUX OIFFERENCE for each f-OPERA 8LE excore channel at least once per hour for the first 24 hours and at least once per 30 minutes thereafter, when the AXIAL FLUX W( OIFFERENCE Monitor Alarm is inoperable. The logged values of the indicated AXIAL FLUX DIFFERENCE shall be assumed to exist during the interval precLding each logging. 4.2.1.2 The indicated AFD shall be considereo outside of its 15% target band when at least 2 0PERABLE excore channels are indicating the AFD to be outside the target band. ' Penalty deviation outside of the 15% target band shall be accumulated on a time basis of: a. One minute penalty deviation.for each one minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and , -
- b. One-half minute penalty deviation for each one minute of POWER OPERATION outside of the target band at THERMAL I"WER levels between 15% and 50% of RATED THERMAL POWER.
4.2.1.3 'The target flux difference of each OPERABLE e.xcoN channel shall be determined by measurement at least once per 92 Effective Full Power Days. The ) provisions of Specification 4.0.4 are not applicable. 4.2.1.4 The target flux difference,shall be updated at least once per 31 Effective Full Power Days by either determining the target flux difference pursuant to 4.2.1.3 above or by linear interpolation between the most recently , measured value and 0 percent at the end of the cycle life. The provisions of (C,~ Specification 4.0.4 are not applicable. ' FARLEY-UNIT 1 3/4 2-2 AMENUMENT NO. 26
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o- m 1 10% T - ~ .______'-- : = _~ 1 _3 ._ __ _ 30% 20% 10% 0 +10% +20% Sb% 3 ! INDICATED AXIAL FLUX OtFFERENCE l l l _ _ _ _ _ _ _ l
INSTRUMENTATION FIRE DETECTION INSTRUMENTATION , hm),htI I g-LIMITING CONDITION FOR OPERATION k UL1 ( i 3.3.3.9 As a minimum, the fire detection instrumentation for each fire detection zone shown in Table 3.3-12 shall be OPERABLE. 1 APPLICABILITY: Whenever equipment protected by the fire detection 1 trstrument is required to be OPERABLE. l l
. 1 ACTION:
I Witn the number of OPERA 8LE fire detection instrument (s) less than the I minimum number OPERABLE requirement of Table 3.3-12: I
- a. Within 1 hour establish a fire watch patrol to inspect the zone (s) with the inoperable f nstrument(s) at least once pcr hour, unless the instrument (s) is located inside the containment, then monitor the containment air temperature at least once per hour at the locations listed in Specification 4.6.1.5.
- b. Restore the inoperable instrument (s) to OPERABLE status within 14 days or prepare and sut,mit a Special Report to the Commission pursuant to Specification 6.9.2 witnin the next 30 days outlir.ing the action taken, the cause of the inoperability and the plans and Schedule for restoring the instrument (s.) to OPERABLE status.
3
- c. Tne provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS '
.53t2'-t "
4.3.3.9.1 Each of the above required fire detection instrumerits which are accessible during plant operation shall te demonstr61.ed OPERABLE at least once per 6 months by performance of a function test which includes subjecting the detector to test aerosol or heat source, as appropriate. Fire detectors which are not accessible during plant operation shall be { demonstrated OPERABLZ by the performance of this functional test during each COLD SHUTDOWN exceeding 24 hours unless perf ormed in the previous 6 months. '. l s 4.3.3.9.2 The NFPA Standard 720 supervised circuits supervision associated with tae detector alarms of each of the above required fire detection instruments shall be demonstrated OPERABLE tt least once per 6 months. FARLEY-UNIT 1 3/4 3-59 AMEN 0 MENT NO. E/,70 b I f
4
,. u ,
TABLE 3.3-12 FIRE DETECTION INSTRUMENTATION (continued) 1xiliary Building - Total- Mininum Room / Smoke 'of Operable Fire Zone Description ~ Elevation Detectors - Sm6ke Detectors 401 Control Roua (above ceiling) 155'-0" 8 4
'401- Control Room (below ceiling) 155'-0" 5 3 440, 455, Clean Toilet, Laundry and 456 Drying Areas 155'-0" 3- 2 462 Non-Radioacti'v e Vent Equip. Rm. 155'-0" 5 3 45S vertical Cable Chase - 155'-0" 3 2 <
456 Vertical Cable Chace 155'-0" 4 2 500' Vertical Cable Chase 168'-2" 7 4 i 416 Ctrl. Rm. Inst. Racks -(above/ ceil.) 155'-0" 9- 5 416 Ctri. Rm. Inst. Racks (below/ ceil.) 155'-0" 6 3 Containment
- 55 Containment Coolers 155'-0" 12/ Fan 6/ Fan 56 Containment 155'-0" 11 6 l Service Water Intake Structure j
. 7 2 A" Pump Room Area 188'-9" 12 6 '
72 A** ' Strainer Bay 167'-0" 18 6 72 B** Switchgear Room - Train B 188'-9" 3 2 72 C** Foyer - Train 8 188'-9" 1 1
'2 0** Foyer - Train A 1F - 3" 1 1 12 E" Switchgear Room - Train A 188'-9" 3 2 (
-73 **- Battery Room - Train B 188'-9" 1 1 ( ;
74 ** Battery Room - Train A 188'-9" 1 1 1 Diesel Generator Building 5o A Switcngear Room - Train A 155'-0" 12 6 56 B Foyer . 155'-0" 4 2 56 C Switchgear Room.- Train B 155'-0" 12 6 71 Hall,4ay 155'-0" 9 5 Diesel Generator Building (Heat Detectors) 57 Diesel Driven Generator 2C 155'-0" 5 3 58 Diesel Driven Generator .1B 155'-0" 5 3 60 Diesel Driven Generator 1C 155'-0" 5 3
-61 Diesel Driven Generator 1-2A 155'-0" 5 3 62 Day Tank Room 2C 155'-0" 1 1 62 Day Tank Room 1B 155'-0" 1 1 65 Day Tank Room 1C 155'-0" 1 1 66 Day Tank Room 1-2A 155'-0" 1 1 ine Fire Detection instruments located within the Containment are not required to be OPERABLE ouring the performance of Type A Containment Leakage Rate Tests.
" inese circuits alarm in the Unit 1 control room area but service both units, anc appear in the Technical Specifications for both units.
FARLEY-UNIT 1 3/4 3-60a AMEN 0 MEN ~ No. 33,70
.. a ,-
TABLE 3.3-12 y FIRE DETECTION INSTRUMENTATION ' (h 9--+ Auxiliary'Buildina Total- hininum
. Room / -Smoke ',of Operable: i Fire Zone' Description Elevation Detectors , Smoke Detectors 128 RHR Heat Exchanger Room 83'-0" 10 5 129 RHR Low Head Pump Room 7 7 '. - 0 " . - 1 1 131 kHR Low Head Pump Room 77'-0" 1 1 160 Hatch Area 100'-0" 3 2 161,162,163 South Corridor 100'-0" 7 4 172 Hallway '100'-0" 3 2 173 Charging /S! Pump Room 100'-0" 2 1 174 Charging /SI Pump Room 100'-0" 2 L 1 175 Hallway 100'-0" 2 1 181 Charging /SI Pump Room 100'-0" 2 1 184 Mechanical Penetration Room 100'-0" 8 '4-185 CCW Heat Exchanger and Pump 100'-0" 13- 7 186 Boric Acid .Trans. PPS 100'-0" 3 2 188 Boric Acid Tanks 121'-0" 2 1 190 MCC Panel Room 100'-0" 4 2 -!
191 Auxiliary Feedwater Pump Room 100'-0" 1 1 192 Auxiliary Feedwater Pump Room 100'-0" 1 1 l 193 AJxiliary Feedwater Pump Room 100'-0" 2 1
-202 Communications Rooms 121'-0" 4:
208,207 2 Corridor 121'-0" 4 2 HalIway 121'-0" 4 2 210,211,228, South Corridor
~
(209 234 121'-0" 7 4 212 - Battery Room 18 121'-0" 213 Battery Service Room 121'-0"
-1 1 l
.214 1 1- -!
Battery Room 1A 121'-0" 1. ; 1 223 Mechanical Penetration Room 121'-0" 9 5 224 DC Switchgear Room 121'-0" 1 1 225 Battery Cnarger*, Room 121'-0" 1 1 226 DC Switchgear Room 121'-0" 1 1 227 West Cable Chase 128'-0" 7 4 229 ~AC Switchgear Room Train B 121'-0" 4 233 '2
. AC Switchgear Room Train B 121'-0" 3 2 241 Main Steam and FDW Valve Room 127'-0" 8 4 244 Mezzanine - Battery Room i 131. ' -0 " . 3 2 245 Mezzanine - Battery Room -
131'-0" 1 3 2 254 Hot Shutdown Panel Room { 121'-0" 1 1 300 West Cable Chase 139'-0" 7 4 312 Corridor 139'-0" 3 2 316,322 Corridor 139'-0" 3 2 31d Cable Spreading Room 139'-0" 8 4 319,339,345 West Corridor 139'-u" 6 3 333 Electrical Penetration Room , 334 139'-0" 1 1 ! Electrical Penetration Room 139'-0" 6 3 l 335 AC Switchgear Room Train A 343 139'-0" 3 2 AC Switchgear Room Train B 139'-0" 3 347 2
,. Electrical Penetration Room 139'-0" 2 1 k FARLEY - UNIT 1 3/4 3-60 AMErcEN Z. 2 3,73
- m. , .
4 ' s EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - T,y> 350*f, MASTER ! LIMITING CONDITION FOR OPERATION i OPERABLE with each subsystem comprised of:3.5.2 Two independen!
- a. !
One OPERABLE centrifugal charging pump, 3 b. One OPERABLE residual heat removal heat exchanger, ' c. One OPERABLE residual heat removal pump, and d. An OPERABLE flow path capable of taking suction from the refueling water storage tank on a safety injection signal and transferring suction to the containment sump during the recirculation phase of operation. i APPLICABILITY: MODES 1, 2 and 3. ACTION: ! a. I With one ECCS subsystem inoperable, restore the inoperable subsystem (, to OPERABLE status within 72 hours or be in at least HOT STANDBY within 6 hours.the next 6 hours and in HOT SHUTDOWN within the following i b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to ! the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances lated actuation cycles to date.of the actuation and the total accumu-i The current value of the usage
. factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
1
/
1 N FARLEY-UNIT 1 e 3/4 5-3 AMENDMENT NO. 26 l !
REACTIVITY CONTRCL SYSTEMS .; FLOW PATHS - OPERATING i LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the'following three boron injection flow paths shall be OPERABLE: . I
- a. The flow path from the boric acid tanks via a boric acid transfer i pump and a charging pump to the Reactor Coolant System.
- b. Two flow paths from the refueling water storage tank via charging pumps to the Reactor Coolant System. .;
. i APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION: With only one of the above required boron injection flow paths to the Reactor Coolant System OPERA 6LE, restore at least two boron injection flow paths to - the Reactor Coolant System to OPERA 8LE status within 72 hours or be in at least HOT STAND 8Y and borated to a SHUT 00WN MARGIN equivalent to at least 1% delta k/k at 200*F within the next 6 hours; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. I SURVEILLANCE REQUIREMENTS 4.1.2.2 At least two.of the above required flow paths shall be demonstrated OPERABLE:
- a. At least once per.7 days by verifying that the temperature of the flow path from the boric acid tanks is greater than or equal to 65 F when it is a required water source and the ambient air temperature of the auxiliary building is less than 65*F.
i 1 .i . b. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct ! position, and .
- c. At least once per 31 days, by verifying that, on recirculation flow, the boric acid transfer pump develops a discharge pressure of 4 '
greater than or equal to 100 psig when the boric acid transfer pump l is required to to be OPERABLE. '
- d. While proceeding to or in COLD SHUTDOWN if not performed in the previous {
12 months by verifying that the flow path required by 3.1.2.2.a delivers ' at least 30 gpm to the Reactor Coolant System. ( $ 1 FARLEY-UNIT 1 3/4 1-8 AMEN 0 MENT NO. 26 l 1
<v ,
1 PLANT SYSTEMS 3/4.7.3 COMPONENT COOLING WATER SYSTEM ~ - 1 LIMITING CONDITION FOR OPERATION 3.7.3 At least two independent component cooling water loops shall be OPERABLE. k APPLICA8ILITY: H0 DES 1, 2, 3 and 4. 1 ACTION: l
' With only one component cooling water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours or be in at least HOT STANOBY within i the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
i i SURVEILLANCE REQUIREMENTS 4.7.3 At least two component cooling water loops shall be demonstrated OPERABLE:
- a. j At least once per 31 days by verifying that each accessible valve ;
(manual, power operated or automatic) in the flow path, servicing safety related equipment that is not locked, sealed, or othemise secured in position, is in its correct position.
~
- b. At least once per 18 months during shutdown, by verifying that each automatic valve servicing safety related equipment actuates to its correct position on a safety injection test signal.
i i
'FARLEY-UNIT 1 3/4 7-11 h..rl0 MENT NO. 26 .
L_______________-___
e o .' e . ,- ; I
'3/4.8 ELECTRICAL POWER SYSTEMS MA-nme h b 3/4.8.1 A.C. SOURCES OPERATING LIMITING CONOITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
- a. Two physically independent circuits from the offsite transmission network to the switchyard and two physically independent circuits from the switchyard to the onsite Class 1E distribution system, an'd
- b. Two separate and independent diesel generator sets (Set A: DG 1-2A and DG-1C, Set B: DG-1B and DG-2C) each with:
- 1. Separatie day tanks containing a m' inimum volume of 900 gallons of fuel for the 4075 kw diesel generators and 700 gallons of fuel for the 2850 kw diesel generators.
- 2. A separate fuel transfer pump for each diesel.
c.- A fuel storage system consisting of four, independent storage tanks each containing a minimum of 25,000 gallons of fuel." APPLICA8ILITY: MODES 1, 2, 3 and 4. ACTION:
- a. With an offsite circuit inoperable, demonstrate the OPERABILITY of
- the remaining offsite A.C. source by performing Surveillance Require-ment 4.8.1.1.1.a within 8 hours and at least once per 24 hours thereafter, and performing Surveillance Requirement 4.8.1.1.2.a, items 1, 2, 3, 4, and 6 on diesel generators 1-2A and IB within 8 hours unless such surveillance has been performed within the pre-vious 7 days. Restore at least two offsite circuits to OPERABLE I status within. 7 days or be in at least HOT STANOBY within the next 6 hours and in COLD SHUT 00WN within the following 30 hours. The provisions of Specification 3.0.4 are not applicable.
b. With one diesel generator set inoperable for reasons other than the yearly scheduled maintenance ** demonstrate the operability of the remaining A.C. sources by performing Surveillance Requirement 4.8.1.1.1.a within 8 hours and at least once per 72 hours thereafter, and per-forming Surveillance Requirement 4.8.1.1.2.a, items 1, 2, 3, 4, and "One operaole fuel storage tank must be available for each required diesel generator.
**If this scheduled maintenance exceeds 10 days, the diesel generator set must be declared inoperable.
FARLEY-UNIT 1 3/4 8-1 l AMEN 0 MENT NO. 26 .
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ELECTRICAL POWER SYSTEMS "O"
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ACTION (Continued) 6 on two* diesel generators within 12 hours.
. generator s' et Restore the diesel to OPERABLE status within 18 days or be in at least HOT followingSHUTDOWN 30 hours. within the next 6 hours and in COLD SH applicable inoperable. if only one of the four diesel generator units isThe c.
With one offsite circuit and one diesel generator set of the above required A.C. electrical power sources inoperable for reasons other than the yearly scheduled maintenance,** demonstrate the OPERABILITY remaining offsite A.C. source by performing Surveillance Require-ment 4.8.1.1.1.a within 2 hours and performing Surveillance Requirement 4.8.1.1.2.a. items 1, 2, 3, 4, and 6 on two* diesel generators within 12 hours. Restore at least one of the inoperable sources to OPERA 8LE status within 24 hours or be in at least HOT STAN0BY within the next 6 hours and in COLD SHUTDOWN with following 30 hours. Restore the other AC power source (offsite circuit or diesel generator set) to OPERABLE status in accordance ; with the provisions of Section 3.8.1.1 Action Statements a or b, as appropriate. d. With two of the above required offsite A.C. circuits inoperable, demonstrate the OPERABILITY of both diesel generator sets by per-forming Surveillance Requirement 4.8.1.1.2.a within 2 hours; unless r the diesel generators are already operating; restore at least one of ( the inoperable offsite sources to OPERABLE status within 24 hours or be.in at least HOT STANOBY within the next 6 hours. With only one } offsite source restored, restore both offsite circuits to OPERABLE status within 7 days from time of initial loss or be in at least HOT j STANOBY following within 30 hours. the next 6 hours and in COLD SHUT 00WN within t e. With both of the above required diesel generator sets inoperable, I demonstrate the OPERABILITY of two offsite A.C. circuits by per-forming Surveillance Requirement 4.8.1.1.1.a within 2 hours and performing Surveillance Requirement 4.8.1.1.2.a. items 1, 2, 3, 4, and 6 on one diesel generator in a diesel set on the other Unit within 8 hours; restore at least one of the inoperable diesel gen-erator sets to OPERABLE status: l "The two ciesel generators chosen to be tested shall verify that at - least at one each train of'LOCA/ shutdown loads is capable of being powered Unit.
**If this scheduled maintenance exceeds 10 days, the diesel generator set must be declared inoperable. -
( FARLEY-UNIT 1 3/4 8-2 AMENOMENT NO. 26 - -
. r
ELECTRICAL POWER SYSTEMS MASTER ACTION / Continued) 1.
.Within 24 hours or be in at least HOT STANDBY within the next 6 :
- DG-2C) hours if are (DGinoperable; 1-2A and DG-2C) or or(DG-18 and DG-1C) or (DG-1C and
- 2. i'
' Within 8 hours or be in at least HOT STANOBY within the next 6 hours if DG 1-2A and DG-18 are inoperable; or 3.
Within 2. hours or be in at least HOT STANDBY within the next 6 1 hours if three or more diesel generators are inoperable. Restore both diesel generator sets to OPERABLE status within 18 days from time of initial loss or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.8.1.1.1 Each of the above required independent circuits between-the offsite transmission network and the onsite Class 1E distribution system shall be: l
- a. , Determined OPERABLE at least once per 7 days by verifying correct breaker alignments, indicated power availability, and- :
b. Demonstrated OPERABLE at least once per 18 months during shutdown by transferring unit power supply from the normal circuit to the alternate circuit. 4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE: '
- a. In accordance with the frequency specified in Table 4.8-1 on a i STAGGERED _ TEST BASIS by:
- 1. Verifying the fuel level in the day tank.
{
- 2. Verifying the fuel level in the fuel storage tanks.
l
\
- 3. Verifying the fuel transfer pump can be started and' transfers fuel from the storage system to the day tank.
4. Verifying the diesel starts and accelerates to at least 900 rpm for the 2850 kw generator and 514 rpm for the 4075 kw generators in less than or equal to 12 seconds. The generator voltage and 1 frequency shall be > 3952 volts and > 57 Hz within 12 seconds after the start siglial and operates Tor 5 minutes. 5. Verifying the generator is synchronized, loaded to 2700-2850 kw for the 2850 kw generator and 3875-4075 kw for the 4075 kw generator and operates for g. eater than or equal to 60 minutes. FARLEY-UNIT 1 3/4 8-3 AMENOMENT NO. 26 ,. _ _ _ _ __----s--
ELECTRICAL p0WER SYSTEMS * '
-___ SURVEILLANCE ' REQUIREMENTS (Continued) MN. -
go %fJ '
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4
- 6. Verifying the diesel generator is power to the associated emergency aligned to provide standby busses. j
. b. i At least once per 92 days by verifying that a sample'of diesel fuel {
from the fuel storage tank obtained in accordance with ASTM-0270-65 is within the acceptable Ifmits specified in Table 1 of ASTM 0975-74 ) when checked for viscosity, water and sediment,
- c. At least once per 18 months by: l 1.
Subjecting the diesel to an inspection and maintenance in accordance with procedures prepared in conjunction with its , manufacturer's recommendations,
- 2. Simulating a loss of offsite power by itself, and:
a) Verifying de energization of the emergency busses and load shedding from the emergency busses. ~ i
)
b) Verifying the diesel starts on the auto start signal, d 4 energizes the emergency busses with permanently connected l
' loads within 12 seconds, energizes the auto-connected shutdown loads through the load sequencer and operates for f i greater than or equal to 5 minutes while its generator is
( l loaded with the shutdown loads. After energization of all j loads, the steady state voltage and frequency of the emergency busses shall be maintained at 4160 420 volts ) and 60 t 1.2 Hz during this test. i
- 3. I Verifying that on a Safety Injection test signal (without loss )
of offsite power) the diesel generator starts on the auto start i signal and operates on standby for greater than or equal to 5 ' minutes. The generator voltage and frequency shall be > 3952 volts and > 57 Hz within 12 seconds after the auto-start signal;
~
the steady state generator voltage and frequency shall be maintained between 4160 2 420 volts and 60 2 1.2 Hz during this test. 4. Simulating a loss of offsite power in conjunction with a Safety Injection test signal, and: a) Verifying de-energization of the emergency busses and load shedding from the , emergency busses. b) Verifying the diesel starts on the auto-start signal, energizes the emergency busses with permanently connected loads within 12 seconds, energizes the auto connected emergency (accident) loads through the load sequencer and FARLEY-UNIT 1 3/4 8-4 e AMENOMENT NO. 26 6 . g
)
EbTRICAL POWER SYSTEMS
/ . SURVEILLANCE REQUIREMENTS (Continued)-
[QQ n. operates for greater than or equal to 5 minutes while its' generator is loaded with the emergency loads. After energization, the steady state voltage and frequency of the emergency busses shall be maintained at 4160 + 420 volts and 60 + 1.2 Hz during this test. ~ c) Verifying that all automatic diesel generator trips, except engine overspeed and generator differential and low lube oil pressure, are automatically bypassed upon loss of voltage on the emergency bus and/or~ a safety injection test signal. 5) Verifying the diesel generator operates for at least 24 hours. During the first two (2) hours of this test, the diesel generators.shall be loaded to 4353 kw for the 4075 kw diesels and 3100 kw for. the 2850 kw diesels and during the remaining 22 hours of the test, the diesel generators shall be loaded to ' greater than or equal to 4075 kw for the 4075 kw diesels and 2850 kw for the 2850 kw diesels. The steady-state generator voltage.and frequency shall be maintained between 4160 + 420 vol ts and 60 + 1.2 Hz during this test. Within 10 minutes after completing this 24-hour test, perform specification 4.8.1.1.2.a.4. l
- 6) \
Verifying that the auto-connected loads to each diesel generator I do not exceed the 2000-hour rating of 4353 kw for the 4075 kw generator and 3100 kw for the 2850 kw generator.
- 7) Verifying the diesel generator's capability to:
{ a) Synchronize with the offsite power source while the i generator 2s loaded with its emergency loads upon a { j simulated restoration of offsite power. I b) Transfer its loads to the offsite power source, and c) Be restored to its standby status. 8) Verifying that with the diesel generators operating in a test node (connected to its bus), a simulated safety injection signal ! overrides the test mode by returning the diesel generator to standby operation. 9) Verifying that the automatic load sequence timer is OPERABLE with each load sequence time within + 10'; of its required value or 0.5 seconds' whichever is greater. 10) Verifying that the following diesel generator lockout features prevent diesel generator starting only when required: a) Oil Temperature High (OTH) FARLEY-UNIT 1 3/4 8-5 AMEN 0HENT HO.45 I h j
gy ELICTR: CAL PO'!".R SYSTEMS SUPMEILLA!!CE REQUIRE!GTS (Continued) b) Coolant Temperature High (CTH) c) Coolant Pressure Low (CPL) d) Crankcase Pressure High (CCPH)
- 11. Verifying the capability to reject a load of greater than or equal to the largest single load associated with that diesel generator (approximately 1000 kw); while maintaining voltage between 3740 and 4580 volts and speed less than or equal to 75!, of the difference between nominal speed and the overspeed trip setpoint.
d. At least once per 10 years or after any modifications which could affect diesel generator interdependence by starting the diesel generators simultaneously, and verifying that the' diesel generators accelerate to at least 900 rpn, for the 2850 kw generator and 514 rpa for the 4075 kw generator, in less than or equal to 12 seconds. e. At least once per 5 years, on a staggered basis, by verifying that the die.sel generator can reject a load of 1200-2400 kw without tripping.
.The diesel generator output breaker (s) must remain closed such that the diesel generator is connected to at least one emergency bus. Ve rify that all fuses and breakers on the energized emergency bus (es) are not tripped.
The generator voltage shall remain within 3330 and a990 volts during and following the load rejection. L FAF. LEY '.N:7 1 3/4 S-6 AMEt!DME!!T t!0.5 7 { L e l
., ~
E1.ECTRICAL POWER SYSTEMS MASTER , SHUTOOWN
-(
f(d6!!$D. D , LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, th6: be OPERABLE: following A.C. electrical power sources shall a. One circuit from the offsite transmission network to the switchyard and from the switchyard to the onsite Class 1E distribution system, and
- b. Diesel generator 1-2A,1C or 18 each with:
1. A day tank containing iminimum volume of 900 gallons of fuel for the 4075 kw diesel generator and 700 gallons of fuel for the 2850 kw diesel generator. . 2. A fuel storage gallons tank of fuel, containing a minimum volume of 25,000 and
- 3. A fuel transfer pump.
APPLICABILITY: MODES 5 and 6. ( I ACTI0N,: With less than the above minimum required A.C. electrical pcwer sources OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until the minimum required A.C. electrical power sources are restored to OPERABLE status.. - SURVEILLANCE REQUIRE.MENTS 4.8.1.2 d The above required A.C. electrical power sources shall be R demonstrated OPERABLE by the performance of each of the Surveillance requirements of 4.8.1.1.1 and 4.8.1.1.2 except for requirement 4.8.1.1.2.a.5. 1 l FARLEY-UNIT 1 3/4 8-6b AMENOMENT NO. 26 - 9
=
l
w .;, . IELECTRICAL POWER SYSTEMS au 3/4.8.-2 ONSITE. POWER DISTRIBUTION SYSTEMS ( A.C. DISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 The following A.C. electrical bus'ses shall be OPERABLE, energized i and aligned to an OPERABLE diesel generator: 4160 volt Emergency Bus F. H and X 4160 volt Emergency Bus G J and L 600 volt Load Centers 0, N. X and R 600 volt Load Centers E, J, L and S 120 volt A.C. Vital Bus A 120 volt A.C. Vital Bus B 120 volt A.C. Vital Bus C 120. volt A.C. Vital Bus D . APPLICABILITY: MODES 1, 2, 3 and 4 ACTION: With less than the above complement of A.C. busses OPERABLE, restore the i inoperable bus to OPERABLE status within 8 hours or be in at least HOT STANOBY ing 30 hours. within the next 6 hours and in COLD SHUTDOWN within the follow-SURVEILLANCE REQUIREMENTS 4.8.2.1 The specified A.C. busses shall be determined OPERABLE and energized at least once per 7 days by verifying correct breaker alignment and, indicated power availability. - FARLEY-UNIT 1 3/4 8-6 c AMEN 0 MENT NO. 26 W g 9
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+ 0
.; 7 s
L gaster COPY 4 ANSWER KEY l l- FARLEY.1&2~ ..
~SRO EXAM REGION II 88/11/14 i
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .______s__._________________.___-__.__________________m.___ ____ _ _ _ _ _ _ _ _ . _ _--
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- 5. . THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 73 THERMODYNAMICS ANSWERS -- FARLEY 1&2 -88/11/14-REGION II MASTER ANSWER 5.01 (1.00) b
)
B. (A longer life time] REFERENCE NUS, Reactor Operation, Neutron Kinetics,. para 5.4, Pages 5.4-1 thru 5.4-4 3.0/3.0 192003K107 ...(KA'S) i ANSWER 5.02 (1.00) A. [Whether the given water-boron mixture absorbs more neutrons than it moderates or moderates more neutrons than it absorbs] REFERENCE NUS, Reactor Operation, Unit 8, Coefficients And Control, Para 8.3 and 8.4, Pages 8.3-1 thru 8.4-4 3.1/3.2 2.9/3.1 2.9/2.9 3.1/3.1 192004K101 192004K103 192004K110 192004K106 ...(KA'S) ANSWER 5.03 (2.00) A. a. [- 5413 pcm] B. a. [- 3428 pcm] (1.0 each] REFERENCE NUS, Reactor Operation, Unit 11, Reactor Core Characteristics, Para 11.1,
.Pages 11.1-1 thru 11.1-6 3.8/3.9 192002K114 ...(KA'S)
I -
- l
, :, i
- 5. THEORY OF-NUCLEAR POWER PLANT OPERATION,-FLUIDS, AND PAGE 74 i' -THERMODYNAMICS 1
i
. ANSWERS -- FARLEY 1&2 -88/11/14-REGION II 1 MASTER 1 4
1
' ANSWER- 5.04 (1.00) COPY 1A. [ Positive due to the faster burnup of Zenon at higher neutron flux levels) -l REFERENCE NUS,: Reactor Operation, Unit 13, Power Operation, Para 13.6, Pages 13.6-1 -thru 13.6-7 3.1/3.1 192006K111 ...(KA'S)
ANSWER 5.05 (1.00) D. [ Direct [Xe] decreases, indirect [Xe] increases, tctal [Xe), increases]- REFERENCE VCS, RT BK:III, RT-12, P 17-23, LO 9. North Anna Training Guide 86.2 Section 4 (Pgs 4.5-4.13), Objective B : Westinghouse Nuclear Training Operations, pp. I-5.66 - 70 KAIP 3.4. NUS, Reactor Operation, Unit 10, Fission Product Poisons, Para 10.4, Pages 10.4-1 thru 10.4-3 3.4/3.4 192006K106 ...(KA'S) ANSWER 5.06 (1.00) D. (Maintaining the rods above the rod insertion limit ensures that acceptable power limits are maintained] REFERENCE Farley Technical Specifications, Para 3/4.1.3, Page B 3/4 1-3 3.4/3.9 ' 192005K115 ...(KA'S) l l I
i
. g .- ,
- 25. . THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND 'PAGE 75 )
THERMODYNAMICS. 1 ( ANSWERS -- FARLEY_1&2 -88/11/14-REGION-II l MASTER 1 ANSWER 5.07 (1.00) . e B. [An immediate rapid rise followed by a gradual increase to a higher i steady state value] J or u no amwer if cod 1,00.1 sinks % k % % w. c h hi'N - REFERENCE i NUS, Reactor Operation, Unit 12, Fuel' Loading.And Startup, Para 12.4, Pages ' 12.4-1 thru 12.4-4 3.9/4.0 192008K103 ...(KA'S) ANSWER 5.08 (1.00) ! C. [0.98] l REFERENCE NUS, Reactor Operation, Unit 12, Fuel Loading And Startup, Para 12.4, Pages 12.4-1 thru 12.4-2 j 3.8/3.8 l 192008K104 ...(KA'S) l ANSWER 5.09 (1.00) D. [Startup rate decreases] REFERENCE
.NUS, Reactor Operation, Unit 13, Power Operation, Para 13.5, Pages 13.5-1 thru 13.5-5 3.3/3.4 192008K117 ...(KA'S) i l
_ _ _ -- _ -__ _ - - - -_ _ _ - - - _ _ _ _ - - - - - - - .l
n:
., g. ~.
g i_ L .5.. THEORY 0F NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE'76 L- . . THERMODYNAMICS f-
. ANSWERS -- FARLEY 1&2. -88/11/14-REGION II l
MASie ANSWER 5.10 (2.00)
' ~
A '. . DECREASE. B.- INCREASE. C. -INCREASE.
.D. DECREASE.
[0.5 each] REFERENCE-TPNP, REV 3, ILP 011-OL, . App A, EO 3; TPNP, Reactor Theory, pp 6-22 thru 6-28.
;NUS, Reactor Operation, Unit 7, Reactor Control, Para 7.4, Pages 7.4-1 thru
-7.4-5 2.9/3.4 3.5/3.7 001000K502 001000K509 ...(KA'S)
-ANSWER 5.11 (2.00)
A. DECREASE. B. INCREASE. C. . INCREASE. D. DECREASE. ! [0.5 each] REFERENCE TPNP, OP 1009.1, ECC. NUS, Reactor Operation, Unit 12, Fuel Loading And Startup, Para 12.5, Pages 3 12.5-1 and 12.5-2 j 3.5/3.6 I 192008K107 ...(KA'S) )1 l 1
t .p I e i-l 15 THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 77<- l THERMODYNAMICS l " ANSWERS -- FARLEY 1&2 -88/11/14-REGION II J MASicd ANSWER, 5.12 ('1. 5 0 ) - A. TRUE.. j B. FALSE. l C. TRUE. [0.5 each] REFERENCE TPNP, REV 2, ILP Subcritical Reactor Theory, EO 5; TPNP, Reactor Theory, pp i 8-27 thru 8-53. NUS, Reactor Operation, Unit 12, Fuel Loading And Startup, Para 12.2, Pages 12.2-1 thru 12.2-3 2.9/3.1 192008K106 ...(KA'S) ANSWER 5.13 (2.00) A. Reactor A B.' Reactor B- [1.0 each] REFERENCE NCRODP-86.2, Section VII, Objective I NUS, Reactor Operation, Unit 12, Fuel Loading and Startup, Para 12.1, Pages 12.1-1 thru 12.1-5 2.7/2.8 3.9/4.0 3.8/3.8 192003K101 192008K103 192008K104 ...(KA'S) l
k 4 j. ji
; 5. : THEORY- OF NUCLEAR POWER PLANT OPERATION, - FLUIDS,. AND ' PAGE 78 THERMODYNAMICS:
l* ANSWERS -- FARLEY 1&2 -88/11/14-REGION II MASTER ANSWER' 5.14 (1.00) A. (Indicated' pressurizer level will be higher than actual pressurizer
-level]
REFERENCE Westinghouse, Thermtl-Hydraulic Principles and Applications to the Pressurized Water Reactor II, 1982, pages.11-27 and 11-28 ' j NUS, Plant Performance, Para 1.2.2, Pages 1.2-3 thru 1.2-11 ' 3.4/3.4 191002K108 ...(KA'S)
. ANSWER. 5.15 -(1.00)
A.- [The. indicated steam flow will be greater than actual steam flow] REFERENCE Steam Generator Water Level Control, OPS-52201B (OLT), Flow, Page 7 3.0/3.3 l 1191002K102 ...(KA'S) ' ANSWER 5.16 (1.00) C. (Steam flow] REFERENCE Westinghouse, Thermal-Hydraulic Principles and Applications to the Pressurized Water Reactor _II, 1982, pages 12-12 and'12-13 NUS, Plant Performance, Para 3.2.4, Pages 3.2-4 thru 3.2-6 3.6/3.8 192008K121 ...(KA'S) '
1 > n 7 l
;5. THEORY-OF~ NUCLEAR POWER. PLANT OPERATION, FLUIDS, AND PAGE 79
- THERMODYNAMICS LANSWERS - -FARLEY.1&2 .-88/11/14-REGION II- ..
MAS d,R fANSWER CDPY
;5.17 - ~(1.00)
B. [DecreaseJin steam pressure during constant >. decay heat] l- REFERENCE. D Westinghouse, Thermal - Hydraulic Principles.and Applications-to~the '
-Pressurized. Water Reactor, 1982, page 14-27 NUS, Plant-Operation, Natural' Circulations, Pages 5 and 8 4.2/4.2 193008K122 ...(KA'S)
ANSWER' 5.18 (1.00) 1 C. [Present SCM is' greater than the expected SCM at normal 100% RTP j operation] REFERENCE TPNP, REV 1,.ILP 0057-OL, App F, Sixth EO; TPNP, Thermo,.pp 2-42 thru 2-50. NUS,' Plant Performance, Natural Circulation, Subcooling, Page 6 3.0/3.2 J3.3/3.4 193003K117 193003K125 ..,(KA'S) ANSWER- 5.19 (1.00)
~
A. -[ Fluid remains a superheated vapor with greater than 100 degrees F of-superheat]
- REFERENCE LTPNP,'REV11, ILP 0057-OL, App F, Fifth EO;--TPNP, ILP 0057-OL, App H, EO 5; TPNP,. Thermo, pp 2-71 thru 2-78. i NUS,. Plant Performance, Para 1.6, Pages 1.6-1 thru 1.6-6 .!
3.3/3.4 f 193003K125 ...(KA'S)
~
-j j
I i l
- 5. THEORY-OF NUCLEAR POWER PLANT' OPERATION,-FLUIDS,-AND .PAGE 80
' THERMODYNAMICS- j ANSWERS -- FARLEY 1&2 -88/11/14-REGION II MASTER ANSWER 5.20 (2.00)
A. DECREASE. B. DECREASE. C.: INCREASE. D. INCREASE. [0.5 each] 4-REFERENCE TPNP,'REV 1, ILP 0057-OL, App I, First EO; TPNP, Thermo. NUS, Plant Performance, Para 8.2, Pages 8.2-1 thru 8.2-4 3.4/3.6 3.6/3.8 193008K105 193008K115 ...(KA'S) i
)
ANSWER 5.21 (1.00) A. ' FALSE. B. TRUE ! [0.5 each] REFERENCE Gansral Physics, HT&FF, pp. 155 & 320 and Subcooled Liquid Density Tables AND N.A. Training Guide NCRODP 83 Section 6, Objective H and Section 3, Objective I-and Section 8, Objective D KAIP 2.4, 2.3,.3.4 . 1 NUS, Plant Performance, Para 1.6, Pages 1.6-1 thru 1.6-6, Para 7.3, Pages 7.3-1 thru 7.3-4 3.4/3.9 2.4/2.5 2.4/2.5 002000K508 193003K102 193004K111 ...(KA'S) i __ _ _ . . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ . _ . . . . _ . . . _ _ _ _ _ - . . _ _ _ _ _ _ _ _ _ _ _ _ . . . _ _ _ _ _ _ _. - _ - - _ _ - _ - _ _ - - - _ _ . - _ _ - - - _ _ _ - _ _ - . _ - - - - _ - _ - _ . . _ _ _ _ - _ - - _ _ _ _ J
1- s j j.:i {. I 'k
!D:.Si THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND' PAGE-81
. THERMODYNAMICS: t ANSWERS'- FARLEY 1&2- -88/11/14-REGION II MASTER 1 ANSWER: .5.22 .(2.00)
A. DECREASE - I B. INCREASE
. C. DECREASE D. DECREASE 1
(0.5 each] ! REFERENCE l '
.NCRODP-86.3, Section II, Objective B NUS, Plant Performance, Para 8.2, Pages 8.2-1 thru 8.2-4 i 3.4/3.6 3 f- 193008K105 ...(KA'S) l ANSWER 5.23 (1.50)
A. TRUE B. FALSE-C. . FALSE [0.5 each] REFERENCE ; NUS, Plant Performance, Para 7.2, Pages 7.2-1 thru 7.2-4 3.5/3.8 2.6/3.1 3.1/3.3 3.1/3.4 015000A101 015000K504 193007K106 193007K108 ...(KA'S) l 4 i
. i .
t a <
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 82 ANSWERS -- FARLEY 1&2 -88/11/14-REGION II )
MASill
?%
1 ANSWER 6.01 (2.00) E < Co. s 3 l A. 1. Resets the step counters on MCB-[0.25] to zero [ 0 . 2 5 -]=
- 2. Resets all urgent failure alarms [0.25] if the conditions have been corrected [0.25]
B. 4. [ Resets the pulse-to-analog convertors in the rod position ] indication system to zero] [0.5] j C. 2. [ Resets the slave cyclers 0-127 counter to zero) [0.5] l REFERENCE I FULL LENGTH ROD CONTROL, OPS-52201E (OLT), OPERATOR LICENSE TRAINING j OBJECTIVE 16, Page 30, 3.5/3.8 1 001000K403 ...(KA'S) j j i ANSWER 6.02 (2.00) A. The digital rod position indication provides indication of the rods ) actual position in the core [0.5]. The demand rod position indication provides indication of the rods demanded position [0.5] ) and assumes that the rod (s) do not stick or drop [0.5]. l l [ Equivalent wording / explanation accepted for full credit] j B. Demand - rod control system (0.25] i Actual - digital rod position indication system [ coils) [0.25] J REFERENCE Digital Rod Position Indication, OPS-52201F (OLT), OPERATOR LICENSE TRAINING OBJECTIVE 10, Pages 16 & 17, 3.4/3.7 014000K406 ...(KA'S) l I
4 6.: PLANT SYSTEMS DESIGN,: CONTROL, AND INSTRUMENTATION PAGE 83 ANSWERS -- FARLEY 1&2 -88/11/14-RI:GION II ANSWER 6.03 (2.00) MASTER ! A. RWST [0.2] Containment sump [0.2] B. 2/4 [0.2] containment pressure transmitters [0.1] > 27 psig [0.1] C. Containment spray pumps [0.2] start [0.2] Containment spray pump to spray header isolation. valves (0.2] [8820A and B] open [0.2) Containment spray additive tank outlet valves [0.2] [8836A and B] open (0.2] REFERENCE Containment Spray And Cooling, OPS-52102C, OPERATOR LICENSE TRAINING OBJECTIVE 4, Pages 9 thru 12, 4.2/4.3 3.9/4.2 4.3/4.5 026000K401 026000A101 026000A301 ...(KA'S)
~.'.' .'
- 6. : PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 84 -
ANSWERS - .FARLEY 1&2 -88/11/14-REGION II MASTER .q l ANSWER- 6.04 (2.50) COPY A.- V503A is electrically interlocked with the high and low pressure steam ! stop' valves of the. feed pump turbines such that V503A.is closed 1 anytime A S/G feed pump steam stop valves are closed. [0.5] [. equivalent wording accepted for full credit] B. Provide an accurate electronit ,w signal [0.25] for calorimetric data [0.25] C. Protect against steam generator . town [0.25] in.the event of a pipe 'j
-rupture [0.25] j D. Return condensate to the condensate Lcorage tank [ dump] [0.25] if I hotwell level > 35 inches [0.25]
E. shut (0.25] l F. . FT 598 [0.25] REFERENCE Condensate and Feedwater, OPS-52104C (OLT), OPERATOR LICENSE TRAINING , OBJECTIVES 3, 5, 9 & 10, Pages 12 thru 29, 3.2/3.4 2.4/2.6 ! 059000K419 056000G007 ...(KA'S) l i i f i
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 85 ANSWERS -- FARLEY 1&2 -88/11/14-REGION II l MASTER ;
ANSWER 6.05 (3. 00) I A. Orifice isolations must be shut to open or close LCV-459 or LCV-460 j [0.25]; fail close on loss-of-air / power [0.25]; auto close with PZR { level = or < 15% [0.25] , 1 B. To open, LCV-459 & 460 must be open [0.125] and pressurizer level > j 15% [0.125]; to auto close, PZR level = or < 15% [0.125], LCV-459 or 460 close (0.125], "T"-signal [0.125] or loss-of-power or air [0.125] ,
)
C. LT-112 modulates LCV-115 [0.15], 71% start divert to RHT [0.15], 81% i full divert to RHT [0.15]; LT-115 at 81% full divert [0.15]; fails ! flow to VCT on loss-of-power / air {0.15] D. To auto close it LCV-115 B, D fully open [0.25], Lo-Lo VCT level (5%) l [0.25], or "S"-signal [0.25] j REFERENCE Chemical And Volume Control, OPS-52101F, OPERATOR LICENSE TRAINING ; OBJECTIVE 4 3.6/3.6 3.6/4.0 3.1/3.1 ! 004000A032 004000K101 004000K106 ...(KA'S) 4 l I i i I
m -l
+ x ' .- .- 1.
j
- 16. PLANT SYSTEMS DESIGN,' CONTROL,.AND. INSTRUMENTATION PAGE 86 ,
-ANSWERS -- FARLEY'l&2- -88/11/14-REGION II
, MASTER ANSWER 6.06 (2.70)
A. 1. SG Lo-Lo level [0.1] of 17%.[0.1] as sensed by 2/3 detectors:(0.1] on 1/3 SGs [0.1] and no LOSP [0,1] d
- 2. Both main feed pumps tripped [0.2] and no'LOSP [0.1]
- 3. An ESF. sequencer signal [0.2]
- 4. An LOSP sequencer signal [0.2]
- 5. AMSAC 2/3-[0.1] <5% Level [0.1] Blocked below C20 [0.1]
B. 1. SG-Lo-Lo level-[0.2] of 17% [0.1] as sensed by 2/3 detectors (0.1] on 2/3 SGs [0.1] ., 2.- Undervoltage signal [0.1] of 70% [0.1] detected-by 1/2 UV relays (0.1] on 2/3 buses [0.1]
-3 . AMSAC 2/3 [0.1] <5% Level [0.1] Blocked below C20 [0.1] !
REFERENCE f Auxiliary Feedwater, OPS-52102H, OPERATOR LICENSE TRAINING OBJECTIVE 1, j 4.5./4.6 061000K402 ...(KA'S). j I I j I
)
a
m- - - _ :9 ,l
.- 6.0 '.. PLANT ~ SYSTEMS _: DESIGN, CONTROL, AND INSTRUMENTATION' PAGE 87 !
ANSWERS:---FARLEY 1&2 -88/11/14-REGION II i MASfER JANSWER 6.07 . (2.50)- R A . -. ' Gripper fully engaged or disengaged [0.5]
- B.- ' Prevent continuing.to pull on a stuck fuel assembly.[0.5]
C. Pushing..down on a fuel assembly with the hoist (0.5]
.t ;D. Only slow-speed can be used when the bottom of the fuel assembly'is '
first entering.the core [0.5] or about to be seated on the lower core-
. plate [0.5]
- REFERENCE Fuel Storage, Handling, Refueling,-And Spent Fuel Cooling, 52108D, OPERATOR-LICENSE. TRAINING OBJECTIVES:4&5, Page 30 1 2.6/3.4 2.5/3.3 2.6/3.3.
-034000K401 034000K402 034000K403 ...(KA'S) -ANSWER 6.08 (2.50) 'l.- Train "A" bypass breaker is closed -2. INPUT ERROR INHIBIT switch in INHIBIT
- 3. MULTIPLEXE'R TEST switch in INHIBIT ;
- 4. LOGIC A switch out of OFF !
- 5. PERMISSIVES' switch out of OFF
' 6. - ' MEMORIES. switch out of OFF ~7. MODE. SELECTOR switch in TEST
[any five (5) @ 0.5-each]
-REFERENCE Rnactor Protection, OPS-52201I (OLT), OPERATOR LICEN_SE TRAINING OBJECTIVE 17, Page 89 3.6/3.6' 012000A403 ...(KA'S) l 4
h l j
,j.
- ca y
- .r . -
.n
' H: 6...; PLANT SYSTEMS DESIGN,-CONTROL,'ANDEINSTRUMENTATION PAGE 88
- q 1
J ANSWERS : -- FARLEY 1& 2 : -88/11/14-REGION,II ' t
-t MASTER ANSWER 6.09 (2.50) ,
'A. ' 1. , main generator protective circuitry '
,, '2. -main transformer protective circuitry
- 3. either of;the unit' auxiliary transformer's protective circuitry -
~;
B. the'S/U transformers -l C. after~15 cycles, an undervoltage or underfrequency condition on the-RCP buses = would' cause the RCPs td trip I [all.0 0.5 each]
' REFERENCE ' Intermediate And Low Voltage AC Distribution, OPS-52103B (OLT), OPERATOR. d LICENSE TRAINING' OBJECTIVE 5, Page 23, 3.1/3.5 062000K410 ...(KA'S) '
i ANSWER 6.10- : (1. 8 0 ) - I A.' From safety injection system accumulator tank?
~B.- CVCS. alternate charging line? ~
l
~
C. From high. head safety. injection, residual heat-removal pumps E.- To residual heat removal pump [0.45 EACH) 1 i REFERENCE Roactor Coolant, OPS-52101A (OLT), OPERATOR LICENSE TRAINING' OBJECTIVE 3, l Figure.12, 3.7/4.0 4.5/4.6 002000K106 002000K108 ...(KA'S) , .. q I i i t l i _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ - . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _____.___________._.__._____..____o
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 89 ANSWERS -- FARLEY 1&2 -88/11/14-REGION II MRSTER ANSWER 6.11 (2.00) h
)
- h. False B. False C. False D. True
[0.5 each] p REFERENCE Pressurizer Pressure And Level Control, OPS-52201H (OLT), OPERATOR LICENSE TRAINING OBJECTIVE 9., Pages 6 thru 12, 3.6/3.7 3.9/4.1 010000K103 010000K101 ...(KA'S) ANSWER 6.12 (3.00) A. 1. Prevent overpressurization of the RHR and CVCS systems
- 2. Prevent depressurization of the RCS
- 3. Prevent overflow and dilution of the RWST
[all 3 0 0.5 each] B. 1. RWST suction valve 8809A(B) [0.25] shut [0.25]
- 2. Charging pump suction valve 8706A(B) [0.25] shut [0.25]
- 3. RCS Press < 402.5 psig [0.25] PT-403 [0.25]
REFERENCE Residual Heat Removal, OPS-52101K (OLT), OPERATOR LICENSE TRAINING OBJECTIVE 3, 3.2/3.5 005000K407 ...(KA'S)
.i. .,
. ' .y-L: .6...~ PLANT SYSTEMS DESIGN,' CONTROL, AND INSTRUMENTATION- - PAGE 90 l
ANSWERS -- FARLEY11&2 -88/11/14-REGION II MASTER. ANSWER 6.13 (0.50) Tho' separation assures that:both trains of'CCW are not sacrificed by one cupply or surge line failing.
.[ equivalent wording accepted-for full credit]
REFERENCE Farley LP OPS-40204A', Objective 6 KAIR 3.3/3.4 l 008000G007 ...(KA'S) 5 I ANSWER. 6.14 '(1.00) LThn valve design limits the plant cooldown rate (0.5] IF any ONE valve Inadvertently sticks open-[0.5) REFERENCE Farley.LP OPS-40201A, Objective 7B KAIR 3.1/3.5 035010K602 ...(KA'S) [
/
. f, I
. . _ _ _ _ _ _ _ ._.-_____m_____-m____ _ _ . _ _ _ _ . _ . __d
!7. : PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 91
~. RADIOLOGICAL CONTROL
' ANSWERS--- FARLEY 1&2L -88/11/14-REGION II MASTER ANSWER 7.01 t (:2. 25 )
- 1. Check the status of the following and-take manual control as necessary: [0.25]
Pressurizer spray valves.[0.25] Pressurizer heaters (0.25] Pressurizer-power operated relief valves [0.25]
- 2. If pressure is decreasing due to a mechanically stuck open spray valve PCV-444C(D) [0.25), then trip the reactor prior to pressure reaching 2000 psig-[0.25], then trip the associated RCP 1A(1B) [0.25] !
; 3 .' If pressurizer pressure is decreasing due to a stuck open PORV [0.25],
then close the associated PORV block valve [0.25]. , 1
'(equivalent ~ wording accepted for full credit] I REFERENCE Pressurizer Pressure And' Level Control, OPS-52201H (OLT), OPERATOR LICENSE ,
TRAINING OBJECTIVE 10, AOP-17.0, Malfunction of RCS Pressure Control System- '! 4.0/3.9 3.7/4.0 3.6/3.9 2.8/4.1 ! 000027A101 000027A215 000027A216 194001A111 ...(KA'S)
\
ANSWER 7.02 (2.00)
-i A. 30 (28-32] rpm / minute B. 26 (22-30] minutes
[1.0 each] REFERENCE Main Turbines And Auxiliaries, OPS-52105A (OLT) OPERATOR LICENSE TRAINING
' OBJECTIVE ~5. Pages 21 - 26, SOP-28.1, 2.6/2.8 045010G010 ...(KA'S)
- _ - _ _ _ _ _ _ _ _ _ _ _ _ = - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ 1
/ e :.
. 3' ,
[7.j L PROCEDURES ' ! NORMAL',; ABNORMAL,' EMERGENCY ANDi PAGE.92'
. RADIOLOGICAL ~ CONTROL'
- ANSWERS:--'FARLEY 1&2 .
-88/11/14-REGION II MASTER 1 ANSWER. . 7. 03'
-(1.50) .' COPY l P : 11 Reduce turbine' load'until vacuum stabilizes
~
9 4 Start circulating water pumps as required.~
~ - 2.
- 3. Start' cooling: tower ~ fans.as required
[all 3:9--0.5 each]
-REFERENCE
'l j
. Condensate And Feedwater, OPS- 52104C (OLT) OPERATOR LICENSE TRAINING 1 OBJECTIVE 12, AOP-8.0, Partial'. Loss of Condenser-Vacuum )
2 .' 6/ 2 . 9 . _ '2. 8/4 .1 1
- 000051G010. 194001A111' ...(KA'S) .
ANSWER '7.04' (1.00) - C.: [ 41 t h'ours ]. REFERENCE 4 Health Physics,. OPS-310 (OLT) OPERATOR LICENSE TRAINING' OBJECTIVE 11', Page 3.3/3.5 23
. 194001K104 ...(KA'S) l i J
d {
. .( .
~7. ' PROCEDURES' : NORMAL' ABNORMAL,.EMERGEi'CY AND , PAGE 93
. RADIOLOGICAL-CONTROL'
, -ANSWERS --'FARLEY;1&2
-88/11/14-REGION II' MASTER -ANSWER 7.05 . (2. 00)
A. False-B. False C. True
'D.- True
[0.5 each) REFERENCE-K Haalth Physics, OPS-310 (OLT) OPERATOR LICENSE TRAINING OBJECTIVE 11, Page 1
- 21. 2.8/3.4 194001K103 ...(KA'S)
' ANSWER 7.06 (1.50) -A. TRUE B.- TRUE f
C. FALSE [0.5 each] j i
- - REFERENCE
- FNP-1-UOP-3.1, POWER OPERATION, Para 3.6, 3.9, & 3.10, Page 3 0400b010 ...(KA'S) 1 1
_ _ - _ _ . ________-____---____-__--___-_-__-__-__.__--a
e-:: . , k;
- i . ::7 L PROCEDURES'- NORMAL, ABNORMAL, EMERGENCY AND PAGE 94
- RADIOLOGICAL. CONTROL
-ANSWERS:--!FARLEY 1&2 -88/11/14-REGION II MASTER
- 7i07
' ANSWER (2.00). !
': 1.
'When RCS pressure < 2315 psig [0.4], verify PZR PORVs [PCV-444B /-
. 445A]. closed-['O.4]
i
- 2. Verify all letdown line orifice isolation valves [HV-8149A, B, C]
L closed-[0.4] or letdown line isolation valves [LCV-459 / 460] closed _[0.4] Verify excess letdown lines [HV-8153 and 8154] isolated [0.4]
~
~ 3. .
REFERENCE -
- FNP-1-ECP-0.0, Loss Of All.AC Power, Step'3, PAGE 3/33, 4.1/4.3 000055G010 ...(KA'S)
ANSWER 7.08. (1.00) Verify SI actuated in accordance with foldout page step 2 , REFERENCE
- FNP-1-ESP-0.1, Reactor Trip Response, Foldout Page 32/32, Steps 2.1 & 6.1
-3.1/4.1 103000G015 ...(KA'S)
E_sm--._u2mmuz-m2._2_w_ 2 .- _ . _ _ _ _ . . . .
. y; 4-
- 1. :7. 7ROCEDURES . - NORMAL, ABNORMAL, ' EMERGENCY AND PAGE 95 j RADIOLOGICAL CONTROL
~ ANSWERS --LFARLEY,1&2 -88/11/14--REGION II l
MASTER ANSWER- 7.09 (2. 7 5)' i A.= 1. 200'
~2. .20
[0.25 each]
.odb (.0'53)
B. ' Align charging. pump ' suction to RWST 40 36t - RWST TO CHG PMP HDR [LCV-115B & D]-open NO.25] & VCT OUTLET ISO VLV [LCV-115C & E] closed {-0,-2St then proceed to Step 5 [n ? c: P o 33 C. ' Establish boration via manual emergency borate valve [0.25]. BORIC ACID TO BLENDER [FCV-113A] [0.25] open [0.25] . V-8439 [ manual' emergency borate valve] [0.25] open [0.25] REFERENCE FNP-1-FRP-S.1 . Response to Nuclear Power Generation /ATWT, Page 5/20 3.6/3.3 3.5/3.2. 3.4/3.4 4.5/4.5 000029A102 000029A103 000029A205 000029G010 ...(KA'S)
. ANSWER. 7.10 (3.00)
A.@ h.'E..'I .. 2 I22"'2t:---
,-..n t-] Apd [a, J
- >b '
;- ,, hm e d o* % 5 5"f b "5 B. Site Area Emergency [SAE]
C.- Site Area Emergency [SAE] [1.0 each]
. REFF.RENCE ' ' Emergency Plan Implementing Procedures, OPS-53002, Objective 28, FNP-0-EIP-19, GENERAL EMERGENCY, Para 3.2.4, Page 3, FNP- O-EIP-18, SITE AREA EMERGENCY, Para 3.2.3 & 3.2.10, Pages 2 &3 3.1/4.4 194001A116 ...(KA'S)
7 .. - PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 96 RADIOLOGICAL CONTROL ANSWERS -- FARLEY 1&2 -88/11/14-REGION II I ANSWER 7.11 (3.00) A. SAFETY INJECTION ACTUATED window [0.3] on j BYP & PERMISSIVE panel lit [0.3] MLB1 9-1 or 19-1 lit [0.3] B. Pressurizer pressure low [0.175] 1850 pnig [0.175] 2/3 [0.175] j Steam Line Differential pressure [0.175] 100 psid [0.175] 1 steam line 100 psig less than other two on 2/3 protection sets [0.175] Low Steam Line pressure [0.175] 585 psig [0.175] 2/3 [0.175] Containment Presssure High [0.175] 4 psig [0.175] 2/3 [0.175] REFERENCE FNP-1-EEP-0, Reactor Trip or Safety Injection, Pages 4/68 and 6/68 4.2/4.1 4.1/4.3 000007G010 000007G011 ...(KA'S) l
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 97 RADIOLOGICAL CONTROL ANSWERS -- FARLEY 1&2 -88/11/14-REGION II COPY ANSWER 7.12 (3.00)
A. 1. Attempt to stop rod motion [0.25] by placing the Bank Selector switch in AUTO or MANUAL [0.25]
- 2. If rod motion does not stop [0.25], THEN trip the reactor [0.25]
B. 1. a. Power range high negative flux rate trip [0.25] may occur [0.25]
- b. Rods will step out in AUTO [0.25] due to Tavg - Tref mismatch [0.25]
- 2. a. Place rod control [0.25] in MANUAL [0.25]
- b. Place turbine [0.25] on HOLD [0.25]
REFERENCE FNP-1-AOP-19.0, MALFUNCTION OF ROD CONTROL SYSTEM, Para 2.3, Page 1, Para 5.2 & 5.3, Page 3 4.4/4.6 3.6/3.8 3.9/3.8 3.9/4.0 000001A205 000003G011 000003G010 000001G010 ...(KA'S)
-7. PROCEDURES - NOhMAL, ABNORMAL, EMERGENCY AND PAGE $8 RADIOLOGICAL CON'13OL ANSWERS -~ FARLEY 1&2 -88/11/14-REGION II MASTER !
ANSWER 7.13 (1.00) [Any'TWO (2) of the following THREE (3) conditions - each worth 0.5 points] 1)_ Action is needed immediately to protect the public health and safety.
- 2) No action consistent with the Technical Specifications (which could) provide adequate or equivalent protection is immediately apparent.-
- 3) The action is approved by the Shift Supervisor, Operations Manager OR the Emergency Director. :
REFERENCE Farley-Administrative Procedure FNP-0-AP-6, Procedure Adherence" AND ! FNP-O-AP-16, " Conduct of Operations - Operations Group", Section 5.0,
" Plant. Operating Procedures" AND (Antonym / Contrast) Operations Memorandum 84-06 KAIR 2.5/3.4 4.1/3.9 194001A103 194001A102 ...(KA'S) i l
l
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 99 RADIOLOGICAL CONTROL ANSWERS -- FARLEY 1&2 -88/11/14-REGION II r
,g pa h '8 % h mS4 ANSiiER COPY 7.14 (2.00)
[Any FOUR (4) of the following FIVE (5) actions - each worth 0.5 points] a) Shutdown turbine as expeditiously as possible - IF the Shift Supervisor deems it necessary, TRIP the turbine b) Shutdown reactor as expeditiously as possible - IF the Shift j Supervisor deems it necessary, TRIP the reactor c) Verify reactor trip bkrs/ bypass bkrs are open at the reactor trip breaker cabinet i d) Verify main turbine tripped at the main turbine governor end pedestal o) Announce (via PA system) " Fire in the Control Room: Shifting control to the Hot Shutdown Panel." j REFERENCE Farley Abnormal Op Procedure FNP-1-AOP-28.2, " Fire In The Control Room"; Farley LP OPS-40204G, Objective 1 AND " Design Bases" Hot Shutdown-Panel AND AOP Qual Reg'mt (AOP-28.2) KAIR 3.3/4.1 4.1/4.5 4.2/4.5 000067K304 000068K312 000068K318 ...(KA'S) 1
i ~ 7$ PROCEDURES - NORMAL, ABNORMAL, EMERGENCY.'AND PAGE %100 RADIOLOGICAL CONTROL
.]
l ANSWERS -- FARLEY 1&2 -88/11/14-REGION II HASTER ANSWER- 7.15 (2.00) a). . Start Boric Acid Transfer Pump 1A or 1B b) .Open Emergency borate to. Charging Pump Valve c). Verify ONE (1) Charging Pump running d) Verify Charging Pump Suction Header Isolation Valves are open (0.5 each]' REFERENCE .-Farley LP OPS-52533A; Qual Regm't Element #CEO-766; FRP-S.1, " Response To , Nuclear Power-Generation /ATWT" AND FNP-1-AOP-27.O, " Emergency Boration" ' KAIR 4.2/4.4 000024K302 ...(KA'S)
1 E 8.. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE %101 h t
-ANSWERS -- FARLEY 1&2 -88/11/14-REGION II r
MASTER ! 1 "2 ANSWER 8.01 (1.50) iJ0,PY 4
- 1. Reason (0.25] recorded in the narrative section (0.25] j L 2. Reported [0.25] to the Shift Foreman (0.25]
- 3. Readings taken as soon as possible (0.25] with the. time noted [0.25]
. REFERENCE FNP-0-SOP-0, GENERAL INSTRUCTIONS TO OPERATIONS PERSONNEL, Para 2.0, Page 1
'3.4/3.4 ,
194001A106 ...(KA'S) ' ANSWER 8.02 (1.00) B. (For operating locked valves in containment and when personnel are in containment, a Shift Foreman may have a key delivered to an individual at the containment entrance] REFERENCE FNP-O-SOP-0, GENERAL INSTRUCTIONS TO OPERATIONS PERSONNEL, Para 7.0, .Pages ! 4 thru 7 3.6/3.7 194001K101 ...(KA'S)
-ANSWER 8.03 (1.00)
C. (DO NOT use the manual operator to force the valve any further'against its seat than the motor operator will drive it. The motor.may not be able to drive the valve off the seat without damaging the operator) REFERENCE FNP-O-SOP-0, GENERAL INSTRUCTIONS TO OPERATIONS PERSONNEL, Para 9.0, Pages 8&9 3.3/3.7 191001K106 ...(KA'S) l 1 I l
[8 . - ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE.%102 ANSWERS -- FARLEY 1&2 -88/11/14-REGION II
' MASTER ;
ANSWER 8.04 (1.00) FNP-0-SOP-O [ GENERAL INSTRUCTIONS TO OPERATIONS PERSONNEL] < Locked Valve Checkout Sheets 1 l STP-64.0, Safeguard System Locked Valve Verification l STP-64.1, Non-Safeguard System Locked Valve Verification [any accepted for full credit] REFERENCE FNP-O-SOP-O, GENERAL INSTRUCTIONS TO OPERATIONS PERSONNEL, Table 2 l 3.1/3.4 194001K105 ...(KA'S) i ANSWER 8.05 (1.50) l 1
- 1. Plant. status ;
i
- 2. Radiological hazards (both 0 0.75 each] l REFERENCE FNP-O-EIP-9, RADIATION EXPOSURE ESTIMATION AND CLASSIFICATION OF !
EMERGENCIES, Para 3.1, Page 1, EMERGENCY PLAN IMPLEMENTING PROCEDURES, OPS 53002, OBJECTIVE 29 3.1/4.4 194001A116 ...(KA'S) -i i ( _- - - ~ . . - - - - - _ . - _ - - . . . . _ _ _ _ _ . _ . _ - - _ - - - _ - . - - _ _ _ . - . . _ - - _ - - - _ - _ _ _ _ - - - _ - _ - _ - . . _ _ . . . - - - - - . _ . _ - _ . , - - - _ _ - . _ _ _ - _ - - . - - _ _ . - - _ _ _ - - - - . _ . _ _ _ _ . _ _ _ -
- . ,7
.84 ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE %103
-l ANSWERS -- FARLEY'l&2' -88/11/14-REGION II-i WIASTER
~. ANSWER 8.06 (2.00)
Restore:LobpA.Tavg=or>541F [0.5] within 15 minutes [0.5] or be in hot standby.[0.5).within the next 15 minutes (0.5] REFERENCE
-Farley Technical Specification 3.1.1.4 3.7/4.1 001010G005 ...(KA'S)
ANSWER 8.07 (1.50) A. '2300 [0.25] - 2500 ppm [0.'25] B . ~. . Restore to 35F or greater [0.5] within 1 hour [0.5] REFERENCE Tachnical' Specification 3.1.2.6, Borated Water Sources - Operating ' 3.3/3.8 004020G005 ....(KA<S) I ANSWER 8.08 (1.50) A. . Night order entries B. Control Room i (0.75 each) REFERENCE Operations Memorandum, Operations Standing Policy Book, ; l Datsd-February 14, 1986 2.5/3.4 o 194001A103 ...(KA'S) 1 l l;
7
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE %104 I
ANSWERS -- FARLEY 1&2 -88/11/14-REGION II l MASTER ANSWER- 8.09- (0.75) 30 seconds
)
REFERENCE Oparations Standing Policy Book, Entry' dated 3-18-87 2.9/3.9 194001A110 ...(KA'S) ANSWER 8.10 (1.00)
- 1. During daily time checks 2.. When chart speed is changed I 38 DJhen r ece < der chcirb e refrued.
ogsyp(0.5 each] REFERENCE Operations Standing Policy Book, Entry dated 12/28/88 3.4/3.4 194001A106 ...(KA'S) I ANSWER 8.11 (1.50) ,
- 1. - Shift Supervisor
- 2. Shift Foreman i
3'. Plant Operator
.[0.5 each]
REFERENCE l
.FNP-0-AP-14, SAFETY CLEARANCE AND TAGGING, Para 4.7, Page 3 !
3.7/4.1 194001K102 ...(KA'S) L l'
_; l
-8. ADMINISTRATIVE' PROCEDURES, CONDITIONS, AND LIMITATIONS. PAGE %105 ANSWERS'- -FARLEY'l&2: -88/11/14-REGION II
' MASTER l ANSWER" 8.12 (1.50) !
- 1. Operations Manager (0.25] for routine administrative functions [0.25]
- 2. On-Call Operations Manager-[0.25] for overall implementation of plant ~l operations during normal operations (0.25]
- 3. Emergency Director (0.25] during events-that require implementation of i the Emergency Implementing Procedures (0.25] j REFERENCE FNP-O-AP-16, CONDUCT OF OPERATION - OPERATIONS GROUP, Para 3.1.7, Page 7 3.6/3.8 194001A105 ...(KA'S) ;
i _m. _ - _ - . - . _
e o
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE %106 ANSWERS -- FARLEY 1&2 -
1./14-REGION II e ffASTER ANSWER 8.13 (3.00) J0PY
- 1. In direct charge of unit operation during startup, power operation, and shutdown
- 2. Supervise the Shift Foreman Operating and Shift Foreman Inspecting to !
ensure proper performance of their assigned duties
- 3. Supervise the Unit Operator and Operator-At-The-Controls to ensure proper performance of their assigned duties
- 4. Ensure shift operations are conducted in accordance with plant procedures and operating licenses
- 5. Approve the removal of equipment and systems from service for maintenance, testing or operational activities
- 6. Coordinate the activities of the Operations, C&HP, Maintenance. I&C, and Security groups to accomplish the operating objectives for his shift
- 7. Notify higher management authority as required by "OFFSITE NOTIFICATION", FNP-O-EIP-26
- 8. Determine the circumstances, analyze the cause, determine that operations can proceed safely and obtain approval from the Emergency Director before the reactor is returned to power after a trip or substantial unscheduled or unexplained power excursion
- 9. Responsible for supervising the On-Shift Operations Group's functions required by the Emergency Plan, the Health Physics Manual, the Security Plan and their respective implementing procedures
- 10. Act as Emergency Director until relieved by higher management authority
[ equivalent wording accepted for full credit] [6 @ 0.5 each] l REFERENCE j FNP-0-AP-16, CONDUCT OF OPERATION - OPERATIONS GROUP, Para 3.1.7.2, Pages 9 & 10 2.8/4.1 3.1/4.1 194001A111 194001A112 ...(KA'S) _ - _ - - _ _ _ i
V- e'
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS,'AND LIMITATIONS. PAGE.%107 2 ANSWERS -- FARLEY 1&2 -88/11/14-REGION II- j hlASTER . I ANSWER 8.14 (2.50)
.A. Once per hour for the first 24 hours (0.25] and at least once.per 30 minutes thereafter [0.25]
B. One (1). hour [60 minutes] [0.5] > C. Indefinitely [0.5] I D.- Reduce thermal power to < 50% [0.2] within 30 minutes-[0.2] and reduce a the Power Range Neutron Flux-High Trip Sotpoints [0.2] to less than or equal to 55% of rated thermal power (0.2] within the next l 4 hours [0.2] i [ Technical Specification 3.2.1 action a.2.a)2) accepted for full- , o r- ke% Afr) h wikn V'far 1.bl I+its in nia.M)<bd REFERENCE - Technical' Specification 3.2.1, AXIAL FLUX DIFFERENCE 3.7/4.1' 001000G005 ...(KA'S) _l l ANSWER 8.15 (2.00)
.7f ;
A. Within 1 hour .{AS-] establish a fire watch patrol to inspect i Fire Zone 191 [3MJ1 at least once per hour JArJ] I
#75
[ Technical Specification 3.3.3.9 action a. accepted for full credit] B. No action is required [0.5] REFERENCE Technical Specification 3.3.3.9, FIRE DETECTION INSTRUMENTATION 3.0/3.6 3.5/4.2 ' 086000G005 194001K116 ...(KA'S) l
c l
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE %108 l
ANSWERS -- FARLEY 1&2 -88/11/14-REGION II MASTER l ANSWER 8.16 (3.00) Within 1 hour [0.5) initiate action to be in [0.25] at least HOT STANDBY [0.5] within the next 6 hours [0.25] at least HOT SHUTDOWN [0.5] within the following 6 hours-[0.25], and at least COLD SHUTDOWN [0.5] within the subsequent 24 hours [0.25] FOLLOWING ALTERNATE ANSWER ALSO ACCEPTED Within 1 hour [0.75], initiate action [0.75] to be in at least hot standby [0.75] within the next 6 hours [0.75] REFERENCE FARLEY - UNIT 1 Technical Specification 3.0.3 '3.5/4.2 006000G005 ...(KA'S) ANSWER 8.17 (1.75) The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator [0.25] will blowdown [0.25] in the event of a steam line rupture [0.25]. This restriction is required to
- 1) minimize the positive reactivity effects [0.25] of the Reactor Coolant System cooldown associated with the blowdown [0.25], and
- 2) limit the pressure rise within containment [0.25] in the event the steam line rupture occurs within containment (0.25].
[cquivalent wording accepted for full credit] REFERENCE FARLEY - UNIT 1 Technical Specification BASES 3.7.1.5, MAIN STEAM LINE ' ISOLATION VALVES 2.6/3.8 000040G004 ...(KA'S) {
?
o
-<' ,4 _
i '8., ADMINISTRATIV
E. PROCEDURE
S, CONDITIONS,'AND-LIMITATIONS: PAGE %109 c:. ANSWERS - ' FARLEY :1&2 -88/11/14-REGION II MASTER ,
'i ANSWER- 8.18 -(1.00)
.Tho.following assumptier.s must be expressed / mentioned:
L a)' No deviation from the original INTENT of the procedure was taken 'I
'b) Approval of the change was apparently made by TWO (2) members of the !
plant. staff, one of which is a licensed SRO ' [0.5 each] REFERENCE -- FNP Technical Specifications, Sec. SIX (6), Pgs 6-12, 6-13 KAIR'4.1/3.9 .
)
,194 001A102 - ...(KA'S) :i ANSWER 8.19 '( 1. 0 0 ) '
l a) REFERENCE FNP-0-AP-16, " Conduct of Operation - Operations Group"; 10_ CFR 50. 54 (i) , 10 CFR 55.4 and 10 CFR 55.14 AND FNP Operations Memorandum 81-02 KAIR 2.5/3.5
-000024K302 ...(KA'S)
I t I
-_ _ __ _____ __ m =\}}