ML20140B900
| ML20140B900 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 03/13/1986 |
| From: | Bill Dean, Wilson B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20140B898 | List: |
| References | |
| 50-250-OL-86-01, 50-250-OL-86-1, NUDOCS 8603250022 | |
| Download: ML20140B900 (115) | |
Text
{{#Wiki_filter:- _ _ _ . _ _ - .. ,o ENCLOSURE 1 EXAMINATION REPORT 250/0L-86-01 Facility Licensee: Florida Power and Light Company P. O. Box 14000 Juno Beach, FL 33408 Facility Name: Turkey Point Facility Docket No. 50-250 Written examinations were administered at Redlan6 Country Club near Homestead, Flcrida. The oral examinations were administered at Turkey Point Nuclear Plant nea'r Homestead, Florida. Chief Examiner: Auu .
.bw' ,'b ih%
WilfiamM. Dean ' Date Signed Approved by: Anu_k. m I3[96 8 9dce A. Wilson, Section Chief Date' Signed Summary: Examinations on February 3-11, 1986 Written and oral requalification examinations were administered to 4 Reactor Operators (R0s) and 9 Senior Reactor Operators (SR0s). Of the 4 R0s tested, none passed the written examination, while three passed the oral examination. Five of the SR0s passed the written examination while six passed the oral examination. l l l l l l 8603250022 860314
'PDR ADOCM 05000250 PDR V
~
2 1 REPORT DETAILS l ! 1. Facility Employees Contacted:
- Bill Miller, Training Superintendent,-Nuclear i *Chris Baker, Plant Manager
- John Crockford, Asst. Superintendent Nuclear Operations
,
- Bob Acosta, QA Supervisor l
- John Shepard, Hot License Instructor
!
- Paul Baum, Operations Training Supervisor
- Leo Goebel, Requalification Supervisor
- Attended Exit Meeting i
l 2. Examiners:
- William M. Dean l William G. Douglas i
- Chief Examiner
- 3. Examination Review Meeting I .
- At the conclusion of the written requalification examinations, the examine'rs j provided the training staff, with a copy of the written examinations and answer keys for review. The training staff'.s comments are included as i Enclosure 4 to this report. The NRC resolutions are listed below.
i a. SRO Requalification. Exam (applicable R0 exam questions in parentheses) . (1) Question 5.02: Based on new material provided, correct answer
- . changed to (a).
! (2) Question 5.10: Based on material presented will accept 12 steps as i equivalent to 15 inches, assuming maximum instrument error. (3) Question 6.04(3.02): Due to unavailability of system to operators and lack of training, question
- is. deleted.
(4) Question 6.06: Due to error in training material provided the NRC,
- Part 2'will be deleted. Upender will be accepted
'as.the correct answer for Part 3.
i (5) -Question 6.09: Based on the additional information provided by
- facility, the recommended answer will be accepted.
1 4
- , - ~ - - - - - - ---m, _ ._ . - , . , - . , _ - . . _ , . _ . - . - . _ _ , . . - - ,m. -,.m.. - - - - - -.
3 (6) Question 6.12(2.14): Due to system not being installed yet, part (a) is deleted, but part (b) will. remain as training was conducted on this system during 1985 requalification cycle. (7) Question 6.13: Based on changes to facility and new material provided, part (a)' answer key will be modified as recommended. Part (b) will require the answer " rod drop sensed by RPI system" to replace recommended deletion. (8) Question 6.14(3.13): Recommended additional answer is similar to existing answer key and will be accepted. (9) Question 7.05: Additional answers in new material provided to the NRC will be accepted. (10) Question 7.06(4.07): Based on new material provided to the NRC, additional answer will be acceptea. (11) Question 7.09: Based on the material provided to the NRC, the recommended additional answer will be accepted. Due to vague wo_rding of question, latitude in grading was given for a description of the heat removal process, e.g., natural circulation. (12) Question 7.10: Based on the new drawing provided the NRC, the three additional answers will be accepted as correct answers. (13) Question 7.12(4.12): Facility's recommended answer is a rewording of existing answer key. No change required. (14) Question 7.13: Lettering or numbering of sub-tasks will be accepted. (15) Question 8.03: Tech Spec 3.0 should be one with which senior operator's are closely familiar. Mode 2 can be possible under the given conditions as long as criticality not achieved <522 F (not 422 F listed in facility comment - apparent misprint). Do not agree having only one unit at power precludes use of TS 3.8.5. This Tech Spec is very difficult to use and should be improved to make it easier for the operators. Answer.(a) will be accepted as an additional correct answer based on the confusion inherent in TS 3.8 and since it is a more conservative action that can be supported by existing TS.
4 a (16) Question 8.09: Based on facility map provided, recommended answers 1 will be accepted. t (17) Question 8.12: Correction to psia will be made to answer key. (18) Question 8.13: Based on inoperability of subject equipment and documentation of such provided to the NRC, question will be deleted. i
- b. R0 Requalification Examination t (1) Question 1.08: Recommended answer is similar to existing answer key. Heat source and heat sink with a-
! AT will be accepted in lieu of density difference. (2) Question 1.10: Based on new material provided to the NRC, recommended change to answer key will.be made, (3) Question 1.11: Question asks for changes -to the- SOM
" calculation", not the SDM. Answer key stands.
(4) Question 1.13: Due to typographical error- in formula sheet ' provided, this question will be deleted. (5) Question 2.02: Based on revised material provided to the NRC, the answer key will be changed to reflect recommended answer. (6) Question 2.06: Recommended additional answer will be accepted. Difference noted is not reflected in facility provided system descriptions. (7) Question 3.03: Due to error in existing system description, , recommended answer will be accepted. , (8) Question 3.06: Due to differences in Units not reflected in material provided the NRC, additional answer . j will be accepted. I (9) Question 3.08: Based on new material provided the NRC that i is not reflected in' current system descriptions, recommended changes will be made to answer key. ] (10) . Question 4.04: Differences in facility operating procedures and current . training emphasis should be resolved. Additional answer recommended will be accepted. 4 (11) Question 4.08: Based on new revision to procedure that was not originally provided to NRC, the recommended change . to the answer key will be made. I i
.. . . ~ _ _ _ _ _ . , _ . _ - _ . . - . . . _ . _ _ _ . . . . , . . _ _ , . . . . , . . . - . _ , _ , _ _ .
- - .. - . = . . . .. _ .-
1 ** .. 1 l i 5 2 Modifications to R0/SRO Requalification Examinations Made During Exam l During the written exam, the following questions were deleted after
; operator comments caused examiners to investigate current _ facility material which had not been provided prior to the examination:
I Question 6.02 (2.03) and 6.07(a)
! Post-Grading Modifications Question 2.11 was deleted from the RO examination. A similar question
~
4 on the SRO examination (6.01) was asked in multiple choice format while 2.11 was in a descriptive format. Since the latter required a greater depth of knowledge it was deemed inappropriate to ask at the
- R0 level.
Question 7.03: Item analysis revealed that the overwhelming majority of i candidates answered choice (b). Review of ES-0.3 and ES-0.4 showed , this to be an acceptable answer.
- 4. Exit Meeting At the conclusion of the site visit' the examiners met with representatives of the plant staff to discuss the results of the examination. Those individuals who clearly passed the oral examination were identified.
3 There were three generic weaknesses (greater than 75 percent of candidates I giving incorrect answers to one examination topic) noted during the oral' ! examination. The areas of below normal performance were knowledge of the l Rod Control System, knowledge of Emergency Response Guideline - Background Information, and lack of familiarity with off-normal operating procedures. The . cooperation given to the examiners and the effort to ensure an
- atmosphere in the control room conducive to oral ~ examinations was also noted ,
l. and appreciated. The licensee did not . identify as proprietary any of the material provided to or reviewed by the examiners. ? l l t e _ ____ _ _ - __._ - ____-_-_______ ______ __ _ - - .
p . . . . . . - g *. ENCLOSURE 3 l
, UINTED STATES ; ' i.8.fm meeg'g '
NUCLEAR RE20LATORY COMMISSION E' mEoioN N f 101 MARIETTA STREET. N.W., SUITE 3900 f ATLANTA, OEoAGIA 30333 i U. S. NUCLEAR REGULATORY COMMISSION i SENIOR REACTOR OPERATOR LICENSE EXAMINATION [E'6@ VAL 1 FACILITY: TURKEY POINT 3&4 I REACTOR TYPE: PWR-WEC3 ; 1 DATE ADMINISTERED: 86/02/03 l EXAMINER: DEAN, W M APPLICANT: _________________________ INSTRUCTIONS TO APPLICANT Use separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up sin C#f hours after the examination starts. %#f' 4
% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 18 0 25 0
___1__0 __ ___1__0 ___________ ________
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 18.00 25.00
________ ______ ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 18.00 25.00 PROCEDURES - NORMAL, ABNORMAL, ________ ______ ___________ ________ 7. EMERGENCY AND RADIOLOGICAL CONTROL 18.00 25.00 ADMINISTRATIVE PROCEDURES, ________ ______ ___________ ________ 8. CONDITIONS, AND LIMITATIONS 72.00 100.00 TOTALS FINAL GRADE _________________% All work done on this examination is my own. I have neither I given nor received aid.
~~~~~~~~~~~~~~
hPPL5C UT I5~E5GU TURE
's 5 km mag g gygygg gyaygg k NUCLE AR RE1ULATORY COMMISSION RE000N II I 101 blARIETT A STREET,N.W., Suf7E 2000 ATLANTA, GEOAQtA 30323
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 2 QUESTION 5.01 (1.00)
Which of the following curves (see attached page) representing Xenon concentration is correct for the given power history? QUESTION 5.02 (1.00) When performing a reactor S/U to full power that commenced five hours ofter a trip from full power equilibrium conditions, a 0.5%/ min ramp was used. How would the resulting xenon transient vary if instead a 2%/ min ramp was used?
- a. The xenon dip for the 2%/ min ramp would occur sooner and the magnitude of the dip would be smaller.
- b. The xenon dip for the 2%/ min ramp would occur later and the magnitude of the dip would be smaller.
- c. The xenon dip for the 2%/ min ramp would occur sooner and the magnitude of the dip would be larger.
- d. The xenon dip for the 2%/ min ramp would occur later and the magnitude of the dip would be larger.
QUESTION 5.03 (1.00) Which of the curves on the following page correctly represents the change in the critical boron concentration over core life? QUESTION 5.04 (1.00) Which of the following would cause an inadvertant dilution accident? a) Overfilling a S/G while in hot standby. b) A ReSenerative heat exchanger leak. c) Valving in a demineralizer that was not saturated. d) A VCT Lo-Lo level resulting in the RWST being used for charging. j (xxx** CATEGORY 05 CONTINUED ON HEXT PAGE xxxxx) i i j
f . . _
, 1 . .:
i s. -
, i
. 3D' '
4 h k 8 . M 1
** ene go see gh g,, i 1
< .. s' I
- a. $, ' .- - - - - - - - - _ -
)
e N 9's e,e is too gao 340 l k E G, , , 30 90 40 to neo ago a'ge - i
- i e
g - C. g .
*
- F w w w ,
so go to to too tao no T m g, f .__ _ . . _ _ _ _ - - - - - c. i
. to go 60 80 soo lao sgo ;l t .
TIME (Howns) l 4 g
- e l FI G U R E
, i o .
D
~ -
_ . . . .Q .. < c __ _..____ . _ . _ _ _ - _ _ _ _ . _ .. - . _ . . . . .._ . _ _ . _ _ _ _ _ . . . .. . _ .
...fe.o . . .
. . . . p _ _ _ _ . _ . _ . . . _ . . . ;
2 ~
; . (W0.
3 Q q . .. .. . . . . . _ . . . _ _ . . . _ _ . ..
<l
.j . _p . . . . . _ . . .... _ ..__ . _ _ p . _ _ _ _ _ . . . . . _ . .
._ . __ . g .s 1 l
l I:
- i. -
j '. i .
.s t . ,
susaaf (d@lhtu) %uwut C&w/Hw) i a. s- to . I_f. . . . . .. 5- . .. .e. o
. . e s-Ji
. { . . _ . .. _ _ _ . , F b
1oon {
. tooc - _ ... ...
4 k .. g) ..
- t. O"9 . e cos -
y s _. 0 -
'\ l
. . j l' Sowve (Cuv/m) Bow & (4WD/Mru) w 'f 1 , r ea ir r l t, i
f~ '
; & UNITIO STATES
}tg*
+ "8! q{9,, NUCLEAR REQULATORY COMMISSION
- o R4040N N 191 MAR 8ETTA STREET,N.W SUITE 2900
{o . f ATLANTA, GEORGIA 30333
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3 QUESTION 5.05 (1.00)
Initially, one centrifugal charging pump is in operation when a second contrifugal charging pump in parallel with the first pump is also put into cperation. Which statement-below correctly describes the effect on system volumetric flow rate and system head loss? a) Higher flow rate, higher head loss b) Same flow rater higher head loss c) Higher flow rater same head loss d) Same flow rater same head loss GUESTION 5.06 -(2.50) The plant is operating at 30% power, turbine in AUTO (IMP IN), when loop 41 reactor coolant pump trips. Assuming a reactor trip does not occure there is no operator action and rod control is in MANUAle indicate whether the following parameters will be HIGHER, LOWER or the SAME at the end of the transient compared to their initial values.
- 1) 42 S/G steam pressure (0 5)
- 2) 43 RCS loop flow (0.5)
- 3) Te in loop 41 (0.5)
- 4) Th in loop 42 (0.5)
- 5) Nuclear Power (0.5)
OUESTION 5.07 ( .50) Does the Latent Heat of Vaporization INCREASE, DECREASE or REMAIN THE SAME os saturation pressure / temperature of water is increased? i (xxxxx CATECORY 05 CONTINUED ON NEXT PAGE xxxxx) i
, ; g aaero UNITED 871.788
',' {o,, NUCLEAR RE1ULATORY COMMISSION
. o REOtoM N 5 'l 101 blARIETTA STREET,N.W SulTE 2000 ATLANTA. 0EORGIA 30323 1
% % , * . . .
- s/ I l
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 4 y
GUESTION 5.08 (1.50) Unit 4 is at 50% power with control rods in MANUAL when the turbine is romped up to 60%. Indicate whether the parameters below will increase,-
- dscrease or remain the same during both the vnitial response (first 30 e coconds of the transient) and after turbine pcwer has stabililzed relative to the initial conditions. (Assume the following: No changes to boron /xer.on Loop transport tine is 10 seconds No operator actions)
NOTE: No answer required where it is already filled in below. Initial Response Steady State a) S/G Pressure NO ANSWER RORD b) Reactor Power NO ANSWER RORD c) Tcold d) Tavs - QUESTION 5.09 (1.50) c) TRUE or FALSE: During cold plant conditionse you would expect the COLD calibrated PZR level instrument to indicate HIGHER than the HOT calibrated level instrument. (0.5) b) Give two different conditions involving the reference leg which could' result in a false high level indication on the PZR level instrument. (1.0) DUESTION 5.10 (2.00) What are the four conditions that Tech Specs say must be met to ensure the Nuclear Enthalpy Rise Hot Channel Factor is maintained within limits during periods between in-core surveillances? l I i 1 (***xx CATEGORY 05 CONTINUED ON NEXT PAGE xxxxx)
,; . f #8 99 UNITED ST'.TES 8 V k NUCLEAR RET.ULATORY COMMISSION RE080 nil 2* 101 tIAAIETTA STREET N.W.,SuffE 2000 ATLANTA. 0EORGIA 30333 -q f
%, . . . . . J
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 5 l
QUESTION 5.11 (2.00) Unit 3 has just restarted following a refueling outage while Unit 4 is naar EOL. Answer the following regarding the differences in plant response between the two units (explain your answers)! o) At a steady power level of 10EE(-8) amps during a startup, equal reactivity additions are made (approximately 100 pen). Which Unit will have the higher steady state startup rate? - b) At 50% power, a control rod (100 pcm) drops. Assuming NO RUNBACK or OPERATOR ACTION, which Unit will have the lower steady state Tavs? QUESTION 5.12 (1.00) What are the two reasons for shifting the SI mode from cold les racirculation to hot les recirculation approximately 24 hours after a LOCA? QUESTION 5.13 (1.00) TS 3.2.2 requires power be reduced to < 75% if a misaligned control rod cannot be aligned within 8 hours. What is the basis for reducing power in this situation? QUESTION 5.14 (1.00) - Operating Procedure 0202.2, Unit Startup, states that during a reactor startup, 'a non-uniform increase in count rate will occur
- when withdrawal of Shutdown Bank A is commenced. What is the reason for this phenomenon and what is the approximate change in count rate that is observed?
(xxxxx END OF CATEGORY 05 xxxxx)
, e UNITED 8YATSS I
~#p$ mee ,qg NUCLEAR REQULATORY COMMISSION e RE040N N D g 101 MARIETTA STREET, N.W., GUITE 2000
- ATLANTA, GEORetA 30323 I
g,*..e*j
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 6 .
QUESTION 6.01 (1.00) According to 10CFR50.46, which of the following is NOT a design criteria of the Emergency Core Cooling System subsystems.
- a. The calculated peak centerline temperature shall not exceed 2000 degrees F.
- b. The maximum cladding oxidation shall not exceed 17% of the total cladding thickness. -
- c. The calculated total amount of hydrogen generated from the cladding reaction with water shall not exceed 1% of the amount that would be generated if all cladding surrounding the fuel reacted.
- d. Calculated changes in core geometry shall be such that the core remains amenable to cooling.
QUE ION 6.02 (1.00) Which the following describes the NORMAL BACKUP source of water to the Fire Pro ion System if electrical power was lost to the Fire Protection System pumps.
- s. The Backup ' e Water pump takes a suction on the Raw Water Tanks and is line o the Fire Protection pump suction.
- b. A spoolpiece is installed b n the Fire Protection System and the Service Water Pump discharge.
- c. The Screen Wash Pump Discharge is connecte a nearby fire hydrant using a hose. -
- d. The Elevated Storage Tank is lined up to provide gravity eed to the Fire Protection Sy tem by opening valve 794 (normally closed).
.-[ .4' (xx*mm CATEGORY 06 CONTINUED ON NEXT PAGE **x**)
' j># #880 9 UNITIO STATSS l
3{# %o NUCLEAR REZULATORY COMMISSION AEGION N { 101 MAAIETTA 8TREET N.W., SulTE 2000 ATLANTA, GEOA04A 30333 1 iv,..... l l
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 7 l QUESTION 6.03 (1.00) ,
l Which of the following correctly describes the actions of the AFW Flow l Centro 11ers AFTER receiving an initiation signal? l
- a. Since the AFW flow control console Hand Indicating Controllers (HICs) are set to a predetermined flow rater the-control valve will RAPIDLY OPEN to this predetermined setting.
- b. Even though there is a pre-set flow rate from the HIC positioners, the flow control valve is initially driven to the FULL OPEN position due to the large error signal between the HIC and the initial O GPM actual flow measured.
- c. Due to a long reset time in the flow control circuitry, the large difference between the HIC setpoint and the O GPM actual flow measured initially, the valve will SLOWLY OPEN to~the setpoint.
- d. Due to a large flow measurement " spike" above the pre-set position on the HIC, the flow control valve will initially STAY SHUT, then as flow stabilizest OPEN SLOWLY to the pre-set position.
bL Y C'J COTIC;: 6;^8 ,, (1.00) Which statement below correctly describes operation of the GAMMA-HETRICS Neutron Flux Monitor in gamma flux fields between 10,000 and 1,000,000 R/hr (ie. high radiation fields).
- a. The monitor is not designed to operate in such high level radiation fields.
- b. The monitor will operate satisfactorily in these radiation levels, but an adjustment should be made to discriminate against the hi3her gamma flux.
- c. The monitor will operate satisfactorily, but the output signal from the detector will not increase. linearly due to lack of voltage saturation in the detector.
- d. The monitor is designed to operate as well in this high a level samma f lu:: as it does in much lower radiation fields.
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
jnameeg UNITED 87.'.788 - kc' NUCLEAR RE2ULATORY COMMISSION REGKHd N { ( 101 M ARIETT A STREET, N.W., SUITE 2000 ATLANTA, GEORGIA 30333 g,.....) < PA E 8 f'__"E!"I_SISIE"SEESIE":_EE"I"SE:_0"B_I"SI"E"E"IeIIEN ) QUESTION 6.05 (1.50) Answer the following questions regarding ESFAS TRUE or FALSE: c) In order to generate a 'P' signale 2 out of 3 Hi Containment Pressure OR 2 out of 3 Hi-Hi Containment Pressure signals are required. b) The S/G Delta P 'S' signal actuates when 2 out of 3 S/G pressure detectors for 1 S/G are 100 Psi GREATER than 2 out of 3 Steam Line Pressure detectors. c) A Reactor Trip in coincidence with a Low Tavs will result in Main Feed Water control valve closure, but the Main Feed Water pumps will still be running if they were operating at the time of the reactor trip. QUESTION 6.06 ( .75) Fill in the blanks in the statement below regarding the spent fuel pool hoist: The hoist and bridge controls are interlocked to prevent raising or lowering the load while the _______ is moving. If the upper limit position switch fails to stop a load lifte the _______ will stop the hoist a few inches higher. The upper limit position switch is set so that the botton nozzle of the fuel assembly will clear the __________. QUESTION 6.07 (1.50) Indicate what automatic actions, if any, occur when high level alarms are received on the following process radiation monitors: U a) R-13 (Plant Vent Air Particulate) Q W M 44 b) R-14 (Plant Vent Gaseous Activity) c) R-19 (S/G Blowdown Liquid Activity) QUESTION 6.08 (2.00) What are the 4 conditions which must be met for the overpressure mitigation system (OMS) status lights to be ON? (xxxxx CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx) l l i l l
-,-m ,v,y
gNITED 8Tf.TES $
. ~ pg >m negg jf ke NUCLEAR REIULATORY COMMISSION nEGoNu
.t '
! l 101 MARIETT A STREET, N.W., SulTE 2000 ATLANTA.GEORelA 30333 iv, d
.....s
- 6. PLANT SYSTEMS ~DESIGNe CONTROL, AND INSTRUMENTATION PAGE 9 OUESTION 6 09 (1.00) I List the 4 sets of ECCS related valves required to mitigate a LOCA which have their control power breakers racked out during critical operations.
QUESTION 6 10 (1.00) List the 4 conditions which will cause the COMMON ALARM on the Remote Post -' Accident H2 Monitoring Panel to actuate (Setpoints not required).
~
OUESTION 6.11 (1.00) o) Preventing steam binding of AFW pumps at Turkey Point is a major concern. What is the potential cause of this steam binding? (.75) ) b) If AFW pump casing temperature is 150 des Fr what action is required ; to reduce the temperature? (Include the frequency of the action) (0.5) l GUESTION 6.12 (1.50) What indication does a control room operator have that a fire damper has actuated? (0.5) b) Explain in detail the design features which allow a fire damper to auto close when required. (1.0) GUESTION 6.13 (2.00) c) Describe the runback process that occurs with the Main Turbine when the OT Delta T setpoint is exceeded? (1.0) b) If the Power Range ' ROD DROP AUTO TURBINE RUNBACK' is bypassed, what conditions must exist and what system will initiate a turbine runback? (1.0) d (xxxxx CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx) i
.' , . ja mweg'g uwTeosTAtas ..
e ', NUCLEAR RE*ULATORY COMMISSION 'j 8' c. naosoN N
- l. ist manmTTA sinsai,u.w., su Ts tooo g ATLANTA.0EOnelA 3o333
\.....
- 6. PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUMENTATION PAGE 10 QUESTION 6.14 (1.00)
The Alternate Source Transfer Switches associated with the recently installed 120 VAC inverters have key locks to prevent 2 switches of the come channel being selected to ALTERNATE at the same time. What are the the purposes behind this administrative key control? QUESTION 6.15 ( .75) Rocently, a backup diesel air compressor and a service air line from Units 1 and 2 have been installed to compensate for a design inadequacy in Unit 3 ond 4's MSIVs. What is this design problem? (xxxxx END OF CATEGORY 06 xxxxx)
. p mee UINTED STATSS g ~
- NUCLEAR RE!ULATORY COMMISSION
.Y e RE040N N <
i ( 101 R$AA8ETTA STREET.N.W., SUITE 1900 ATLANTA, GEORGIA 30333
\, . . . . . )
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 11 '
~~~~
Rd65 E66 6dE~C6HTR6E~~~~~~~~~~~~~~~~~~~~~~~~ QUESTION 7.01 (1.00) The Response Not Obtained for the first immediate action of E0P-FR-S.1
' Response to Nuclear Power Generation / ATHS' is to manually trip the reactor. If the reactor will not tripe thent
- a. Place rods in Manual and insert them into the core.
- b. Trip the turbine and verify steam dumps open.
- c. Emergency borate the RCS.
- d. Dispatch operator to locally trip reactor.
QUESTION 7.02 (1.00) Which of the following reasons correctly describes the basis for allowing RCP restart in E0P-FR-C.1 ' Response to Inadequate Core Cooling'.
- a. Helps to mix the SI flow to protect reactor vessel from cold water. .
- b. Once subcooling is established, restarting the RCPs helps to collapse voids that may have formed in the reactor vessel head.
- c. Allows restoration of PZR pressure control using normal sprays.
- d. Provides for cooling of the core when secondary depressurization does not alleviate inadequate core cooling.
(xxxxx CATEGORY 07 CONTINUE 0 ON NEXT PAGE xxxxx) l l
gneog UNITED STATSS 81 D f k e-NUCLEAR REIULATORY COMMISSION REGION N
)
3 ( 101 MARIETTA STREET, N.W., GUITE 2000 ATLANTA, GEORGIA 30323
%, . . . . 4 l
l I
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 12
--- sA5i5E55iEAE 55 sis 5E------------------------
l l QUESTION 7.03 (1.00) , l E0P-ES-0.3r ' Natural Circulation Cooldown with Steam Voids with RVLMS'r ~ l allows concurrent cooldown and depressurization of the RCS. Which state-ocnt below correctly describes how E0P-ES-0.4, ' Natural Circulation Cool- ; down with Steam Voids without RVLMS", compares with the actions in ES-0.3?
- a. The procedures are identical in that they allow concurrent cooldown and depressurization, except ES-0.4 is done at a slower rate.
- b. The procedures are identical in that they allow concurrent cooling down and depressurization, except ES-0.4 has you monitor PZR level vice RVLMS for void formation.
- c. ES-0.4 uses auxiliary spray as the primary method of depressur-izing, while ES-0.3 uses a PZR PORV as the primary method.
- d. Temperature and pressure are decreased in specified increments on an alternating basis in ES-0.4 vice concurrently.
QUESTION 7.04 (1.50) . Answer the folowing questions regarding E0P usage TRUE or FALSE: a) If a Function Restoration Procedure (FRP) is entered due to an ORANGE Critical Safety Function (CSF) condition, and a HIGHER priority ORANGE condition is encountered, the original FRP must be completed prior to proceeding to the newly identified FRP. b) Unless.specified, a task need not be fully completed before proceeding to a subsequent step as long as that task is progressing satisfactorily c) If a procedure transition occurse any tasks still in progress from the procedure which was in effect need not be completed. QUESTION 7.05 (1.50) l l List 5 possible alarms (setpoints not required) on the Main Control Board ' that would be indications that an inadvertant dilution were occurring while the Unit was at power. (Assume Rod Control is in MANUAL, NO Rx Trip occurs and NO operator actions are taken to mitigate the dilution) (xxxxx CATEGORY 07 CONTINUED ON NEXT PAGE xxxxx)
. am noe UNITED 8TATRS
, '/ k NUCLEAR REGULATORY COMMISSICN I i
! namow s:
C { 191 HARIETTA STREET, N.W 8UITE 3900 ATLANTA, GEORGIA 30333 -
- 7. PROCEDURES - NORMAle ABNORMAL, EMERGENCY AND PAGE 13
~~~~~~~~~~~~~~~~~~~~~~~~
~~~~RA5iBL55iEAE EbETR6L QUESTION 7.06 (1.00)
List the two conditions (including setpoints) which determine Adverse Ccntainment conditions. CUESTION 7.07 (1.00) List the 4 conditions that must be met in order to perform a startup following a Reactor Trip without completing an ECC. QUESTION 7.08 (1.50) List the three conditions stated in Tech Specs which make a control rod INOPERABLE. , QUESTION 7.09 (1.00) H:w is the RCS cooled during refueling operations with the refueling cavity full, if BOTH RHR pumps fail to operate? QUESTION 7.10 (1.50) Unit 3 is shutdown, 4KV Bus 3A is deenergized, A EDC and 43 Startup transformer are both INOPERABLE. It is required that certain vital loads an Bus 3A be operated. List three methods (including power source and cny interim buses) by which this bus can b e reenergized. QUESTION 7.11 (2.50) List ALL the immediate action sub-steps from E-0, " Reactor Trip or Safety Injection' that allow you to accomplish the following immediate actions: a) Check if SI actuated (1.2) b) Containment Ventilation Isolation (1.2) I i c) Verify AFW pumps running (0.6) (xxxxx CATEGORY 07 CONTINUED ON NEXT PAGE xxxxx)
'l l I-
.' p ase s, gNiTED ST'.TES r.
, .'# g NUCLEAR RE00LATORY COMMISSION jl
[ c, RE000N N O
$ it1 MARIETTA STREET.N.W.,8UffE 2000 q
- ATLANTA, GEORGIA 30333 m...** ,
- 7. PROCEDURES - NORMALe ABNORMALe EMERGENCY AND PAGE 14 ,f
~~~~R I65 E65565L'66 TR6L'~~~~~~~~~~~~~~~~~~~~~~~
I QUESTION 7.12 (1.50) During a small break LOCA (SBLOCA), it is required to trip the RCP if the f trip criteria are net. If forced flow through the core promotes coolinge why are the RCPs tripped. - OUESTION 7.13 (2.00) Answer the following questions ~regarding E0P usage! c) What indication is used in the procedures to denote sub-tasks which must be performed in sequence? (0.5)
~
b) What operator action is required if a " Response not Obtained' , contingency action is required, but CANNOT be successfully completed and further contingency actions do not exist? (0 5) ,, c) What operator actions are required ife during performance of steps in a ORP (Optimal Recovery Procedure), an ORANGE terminus on a CSF Status *' Tree is encountered? (1.0)
)
I
)
(xxxxx END OF CATEGORY 07 xxxxx) 1
- - , we - - - - - , , - - - -. - - -
r
~.
.[ neog UNITED STATES k
NUCLEAR RE00LATORY COMMISSION , o REet0N N { l 101 ISAR8ETTA STREET, N.W., SulTE 2000
- ATLANTA, GEORGIA 30333 1
_I___ _ _____ _ ___________$____ _ _ _$______ _ ___ ___ QUESTION 8.01 (1.00) Which of the following describes the MINIMUM review requirement for cpproving an On the Spot Change (OTSC) to a procedure. (Procedure intent is not changed) a) Plant Manager b) PS-N and RCO c) Plant Nuclear Safety Committee _ , d) Two members of plant management, both having an RO license e) Two members of plant management, one having an SRO license OUESTION 8.02 (1.00) Which of the following statements correctly describes proper status of containment building penetrations during refueling operations? -
- s. Both air lock doors can be OPEN as long as'an individual is stationed to shut one of the doors if conditions require this action. -
1
- b. Penetrations leading from the containment atmosphere to the j outside atmosphere can be OPEN as long as an OPERABLE automatic isolation valve is in place.
- c. The equipuent door is OPEN but capable of being immediately shut.
- d. The Containment Purge system is SECURED and all isolation valves ;
SHUT. l l l l (xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx) l l l 1
= ,' M STATSa ,
' ',cg neog ( NUCLEAR REQULATORY COMMISSION 8 M000N N
{e 101 MAA8ETTA STAEET.N.W 0UITE 2000 y ATLANT A, GE0AGIA 30333 l k*.* . t
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 16 QUESTION 8.03 (1.00)
Using the attached technical specification, which action below would be 1 correct for the following situation? SITUATION! Unit 3 at 50% power, i Unit 4 in Startup mode with Tavs = 410 Deg, C AFW Pump is taken out of corvice due to a surveillance (All other AFW equipment is OPERABLE) . ,
- a. Unit 4 must be cooled down to < 350 degrees within 12 hours. ,
- b. EITHER Unit 3 OR Unit 4 must be shutdown / cooled down to < 350 y degrees within 72 hours.
- c. BOTH Unit 3 and Unit 4 must be shutdown and cooled down < 350 de3rees within 12 hours if C AFW pump cannot be restored within 72 hours. -
- d. Unit 3 must be shutdown and cooled down < 350 degrees within 12 hours if C AFW pump cannot be restored within 72 hours. ,
- e. No action is required.
QUESTION 8.04 (1.00) According to Tech Specs, the secondary activity limit is _____ micro-curies por gram and is based on a ______. -
- a. 1.0, Load Rejection
- b. 1.0, S/G Tube Rupture
- c. .67, S/G Tube Rupture
- d. .67, Load Rejection l
(xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xx***) l I l t _.
. u
- 18 STEAM AND POWER CONVERSION SYSTEM 5 C
. 1
*l Mlicability: Applies to the operating status of the steam and power conversion g
. systems.
. q
- Obloctive: To define conditions of the steam 4311eving capacity and auxillary feedwater system.
,h Specification: 1. When the reactor coolant of a nuclear unit is heated above 5300F, the
. .following conditions must be met:
4
,-. a. TWELVE (12) of its steam generator safety ' valves shall be f
operable (eitcept for testing).
- b. Its condensate storage tank shall contain a minimum of 185,000 gallons of water.
- c. Its main steam stop valves shall be operable and capable of closing in 5 seconds er less.
- d. System piping, interlocks and valves directly associated with the related components in TS 3.8.1 a, b, c shall be operable.
. 2. The lodine-131 activity on the secondary side of a steam generator shall not exceed 0.67 $1/gm.
j . .- , 3. With the reactor coolant system above 3500F, if any of above j , ,
.',. . specifications cannot be met within 48 hours, the reactor shall be C~ ~- -
shutdown and the reactor coolant temperature reduced below 3300F. Wification 3.0.1 applies. I
- 4. The following number of Independent steam generator auxillary feedwater trains and their associated flow paths (steam and water) shall be operable when the reactor coolant is heated above 3500F
i j 1 l 3.8-1 Amendment Nos.H0 and #
fl
,i . ~!
)
kn
, s. Sinale 84uclear Unit Oseration Two hdependent aualliary feedwater trains capable of being powered from an operable steam supply.
- b. Dual Nuclear Unit Oseratten -
Two independent auxiliary feedwater trains and a third pump ~
' capable of being powered from, and supplying water to elther traln.
i h.. ' ' c. It in accordance with Ts 4.10.I, testing is required during start-up of either unit, T5 3.L4.a. er b., as applicable, shan apply for an auxillary feedwater pump, pumps, or associated flow noths (steam ; and water) found to be Inoperable. .;
- 3. During power operation, if any of the conditions 'of 3.8.4 cannot be 7
met, the reactor shaR be shutdown and the reactor coolant temperature reduced below 3308F, .unless one of the following conditions can be mets
- a. For single unit operation with one of the two required independent aux 1Rary feedwater trains inoperable, restore the Inoperable train to operable status within 72 hours or the reactor shall be shutdown and.the reactor coolant teroperature reduced below
- f. 3308F whhin the next 12 hours.
Qi '
- b. Per . dual emit operation, one auxt!!ary feedwater pump and its W piping, valves, and Interlocks may be Inoperable
, , provided two M+p1.; auxillary feedwater trains remain operable for time period not to escoed 72 bours. If the Inoperable pump cannot be made operable within 72 hours, one reactor shall se shutdown and its reactor coolant temperature reduced below 3308F whhin See next 12 hours.
- c. For dial unit operation, with one Independent auxillary feedwater train inoperable in one reactor, the affected reactor shall be 5HUTDOWN and its reactor coolant temperature reduced below 3308F within 72 hours. T5 3.L5.a applies for the single unit still g, in operation.
d
- d. For dial emit operation, with one independent auxillary feedwater train inoperable in both units, one reactor shall be 5HUTDOWN and its reactor coolant temperature reduced below 3500F within 12 hours. TS 3.8.5.a applies for the single unit stillin operation.
i
- 3. 8.? Asrahah H0 ng) 104
. '. i weefassfates
' ' NUCLEAR REQULATORY COMMISSION
. *% nSes0N N
{ f 101 SAA8METTA STREET. N.W., SulTE 2000 ATLANTA, GEORGIA $0828 - J l l
%s* .e*4 .
l i
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 17 GUESTION 8 05 (1.50)
Answer TRUE or FALSE to the following.
- o. Entry into an Operational Mode may be performed even if the conditions for the Limiting Condition for Operation (LCO) are NOT net provided the ACTION requirements are subsequently satisfactorily completed.
- b. If a LCO is NOT net and the ACTION statements are NOT appli-caple, then the Senior Reactor Operator.has the authority to disregard that particular LCO. -
- c. Failure to complete a Surveillance Requirement on operable equipment within the specified time interval (plus any allow-able extension) shall constitute a failure of the component to meet its operability requirements.
QUESTION 8.06 (1.50) . Indicate whether the following situations violate the Technical Specification for Power operations on UNIT 43 , c) 'A' EDG out of service with Unit 3 shutdown and 3A 4KV Bus deenergized. b) *B' EDG eut of service with Unit 3 at power and 4A RHR Pump is out of service c) 'A' EDG out of service with Unit 3 shutdown and 3B HHSI pump out of service. QUESTION 8.07 (1.00) List the two individuals who may authorize equipment clearances. QUESTION 8.08 (1.00) Under what two conditions may the Operator at the Controls (OATC) leave the Surveillance Area (without bein3 relieved of his duties)? (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) l 4
,-. - - - . , , - ~ , _ , - - - - - . . - - - . . _ . ,-. . - . . . ,
' % mutsest;tes NUCLEAR RE2ULATORY COMMISSION 4 ke AssoN u
! l 101 MAA4ETTA STASET, N.W., SutTE 9990 ATLANTA, SE0A04A 30333 0
4 1
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 18 4 QUESTION 8 09 (1.50)
List the three meteorological conditions which would preclude conducting O routine Gaseous Waste release. QUESTION 8.10 (1.50)
-List 5 acceptable methods by which Independent Verification of electrical -
breaker alignments may be accomplished. QUESTION 8.11 (1.50) What are the three responsibilities of the Emergency Coordinator which CANNDT be delegated. QUESTION 8 12 (1.50) , List the DNB related parameters as stated in Tech Specs, and their cotpoints. (Assume normal power operations) diHfESTTOM 9 19 (1.oo) , q(f Where is the access code for the Autodialer retained? (0.4) ph' How is operability of the Autodialer verified prior to being utilized in an Emergency situation? (0.6) GUESTION 8.14 (2.00) List 4 of the 5 general conditions in the facility's 'Re-Entry' emergency procedure under which the Energency Coordinator may authorize entry into en area that has already been evacuated. (xxxxx END OF CATEGORY 08 xxxxx) (xxxxxxx xxxxx END OF EXAMINATION xxxxxxxxxxxxxxx) l l
g,__ out)/(Inorgy h ik (
'V'- o o og s o y,t o 1/2 at 2 -
t = ac t " 1 KE = 1/2 av t
, , gyf , y p)fg g , 3, g ,g ,-at y PE =.agn !
Vf = V, + at w = e/t a = an2/t1/2 = 0.693/t1/2 W = w 2P
~
A= sD, 2 1 1/2#*bnM [(typ,)+(t,)) aE = 931 as - i
. m = V,'Ae -Ex .!
o .= ma .h I
- I ,e j o = aCpat 6 6 = UAa T I
- I,e~"* ;
Pwr = Wfah I = 1,10~"E
- TVL = 1.3/v i 8 ~
P = P,10 "#III HYL = -0.693/v p=Pe/I e t SUR = 26.06/T SCR = 5/(1 - Kgf) SUR = 26 lp + p)/(lieff-p) CR, = 5/(1 - Egf,) SUR = 26e/a* + (s - e)T s CRj (1 - K df1) = CR2 II ~ "eff2) T = (a*/s) + ((a - eFIe] f , M = 1/(1 - Kg f) = CR j/CR, i T = a/(e - s) M = (1 - Kgf,)/(1-Kgfj) - T = (s - e)/(Te) 50M = ( -Kgf)/Kgf a = (Kgy-1)/Kgf = aKgf /K gf s' = 10 seconds . I = 0.1 seconds *I -
.e = ((a*/(T Kgf)]+[sgf (1 / + IT)]
Ijj=Id d Id 2 ,2 gd 2 P = (seV)/(3 x 1010) jj 22 I = eN 2 R/hr = (0.5 CE)/d (meters) R/hr = 6 CE/d 2 gf,,g) . ] Water Parameters _ Miscellaneous Conversions t 1 1 gal. = 8.345 lem. I curie = 3.7 x 1010dps
- I ga; . = 3.78 liters I kg = 2.21 lbm i 1 ft* = 7.48 gal. I ap = 2.54 x 103 8tu/hr .
i Density = 62.4 Ing/ft3 1 nw = 3.41 x 106 Stu/hr Density = 1 g n/ car' lin = 2.54 cm i Heat of vaporization = 970 Btu /lem 'F = 9/5'c + 32 Heat of fusion = 144 Stu/lbe ,
'C = 5/9 ( *F-32) 1 Atm = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-lbf 1 ft. H 2
= 0.4335 lbf/in.
e = 2.718 eh '" 3N5 -] . - u
- - - _ . - - _ . . _ -- . _ _ _ . . _ _ . . - . - - _ . - _ _ _ _ - - _ . . _ _ _ . - . . - _ . . _ _ ~ , - , .
** M*g UNITED STATSS
-(daf k r NUCLEAR RE^ULATORY COMMIS8K)N AsteOn u h
2 101 MAAIETTA STREET.N.W., SulTE 2600 h'4 l
- *, ATLANTA.0E0melA 30333
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 19 ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, HM ANSWER 5.01 (1.00) c REFERENCE Ccaprehensive Nuclear Training Operations (CNTO), pp 4-16/27 001/000; K5.13(3.7/4.0)
ANSWER 5.02 (1.00)
/ C. -A:
REFERENCE CNTO W 5* ~ '~Reactor 001/000; w M Core WK5T 38(3.5/4.1) Book "Rs Control'Men &Section Case %vu4 ' psq r.7.gy ANSWER 5.03 (1.00) b REFERENCE CNTO, ' Reactor Core Control'r pp 5-10 001/010; K5.31(2.3/3.1) ANSWER 5.04 (1.00) c REFERENCE TPT Requal Lesson Plan, Cycle II, Day 1-1985 TPT SD13, 'CVCS*r pp 23
n gugygD STATSS
* , /,pa se ,g NUCLEAR RE00LATORY COMMISSION
.-' e nEneoM u
$ ( 101 MARIETTA STREET N.W., SUfTE 2000 ATLANTA.GEoAGIA 30323 4
'-
- MilCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 20 ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WH 004/020; A2.13(3.4/3.9)
ANSWER 5.05 (1.00) D REFERENCE CNTO, ' Thermal / Hydraulic Principles and Applications, II'e pp 10-45/48 006/050; K5 01(2.9/3.1) ANSWER 5.06 (2.50)
- 1) Lower (Higher Sta Flow >> P sta decreases)
- 2) Higher (Less resistance to flow >> Other RCPs speed up)
- 3) Lower (Less total flow across core >> delta T increases, Tc goes down with rods in manual)
- 4) Higher (as above, delta T increases, Th increases)
- 5) Same (Primary power = secondary load)
REFERENCE NUS, Vol 4. Units 1.3, 3.2 CNTO, ' Thermal / Hydraulic Principles and Applications *r pp 12-15/18 002/000; K5.01(3.1/3.4) ANSWER 5.07 ( .50) d2 crease REFERENCE CNTO, ' Thermal / Hydraulic Principles and Applications, I*, pp 2-58/59 000/027; EK1.03(2.6/2.9)- l l l 1 l
~
*****- uswTeosTATes
/, g NUCLEAR REGULATORY COMMISSION 8 7*
#E000N N 2 i 1st asARETTA STREET,N.W SulTE 2000 ATLANTA, GEORG8A 30323
,, . . . . . )
1
- vucnRY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 21 ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WM ANSWER 5.08 (1.50) c) decreasei (no answer) (+.25 es response) b) (no ans); increase c) decreasei decrease d) decreasel decrease REFERENCE CNTO ' Thermal / Hydraulic Principles II'r pp 12-39-45 039/000; A2.05(3.3/3.6)
ANSWER 5.09 (1.50) o) False (+.5) b) -Post accident heating of Reference Les (+.5 ea)
-Reference Les leakage REFERENCE NRC IE Info Notice 84-70 (4 Sep 1984)
TPT Lesson Plsn for Requal Cycle II-1985 - 011/000; K4.03(2.6/2.9) ANSWER 5.10 (2.00) g f g 3 f,gJ )
- 1) Control rods within + or - 15 inches of group demand position (+.5 ea)
- 2) Proper sequencing and overlap of rod groups and rods move together
- 3) Control rod insertion limits are maintained
- 4) AFD is maintained within limits REFERENCE SON TS B 3/4 2-2 TPT TS B3.2.5 001/0003 K5.46(2.3/3.6) i 1
~
,' to Mo gn TED STATES j
", # uq{g NUCLEAR RE ULATORY COMMISSION
! l' 101 MARIETTA STREET, N.W., SulTE 2000 ATLANTA,0E0AGIA 30333
*tr
'- ~*v OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 22 ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WH CNSWER 5.11 (2.00) c) Unit 4 (+.5) due to a lower Beta coefficient at EOL (+.5) b) Unit 3 (+.5) due to MTC being less negativer so Tavs must decrease come to add + reactivity) (+.5)
REFERENCE CNTO ' Reactor Core Control', pp 3-21 & ' Fundamentals of Nuclear Reactor Physics", pp 7-31 001/0001 K5.49(2.9/3.4) & K5.10(3.9/4.1) CNSWER 5.12 (1.00) remove boric acid that is precipitated on upper core surfaces (+.5) terminate any boiling or steam formation in upper head region (+.5) REFERENCE Wastinghouse PWR Systems Manual, pp 4.2-27 , TPT SD-21, 'ECCS*, pp 26 EPE-011: EK3.13 (3.8/4.2) ANSWER 5.13 (1.00) Ensures design margins to core limits will be maintained (+.75) under both steady-state and anticipated transient conditions (+.25) REFERENCE TPT TS B3.2.2 000/005; EK1.06(2.9/3.8) ANSWER 5.14 (1 00) This is due to " Uncovering' of the sources by that bank. (+.7) causing a ap p r o >:i m a te a l y 1/2 decade increase in count rate (+.3) REFERENCE TPT OP 0202.2, Step 4.2.2 001/000; K1.05(4.5/4.4)
l
** #49 UNITED $70738 l 9'S NUCLEAR REGULATORY COMMISSION
'. % RE000N N 3 1et MAA6ETTASTRttT,N.W., SUfTE 3000 ATLANTA.SECAGIA 30333
*fr
**ee*
I
- 6. PLANT SYSTEMS DESIGNe CONTROL, AND INSTRUMENTATION PAGE 23 "
ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, WH l ANSWER 6.01 (1.00) ; o ! REFERENCE 10CFR50.46(b) ) FNP, SD, 'ECCS', pp 5 - -- NA NCRODP 91.1, 'ESF-ECCS' ) 006/050; PWG-4(4.2/4.3) l ANSWER 6.02 .00) d REFEREN TPT 53 ' Service and ire Water', pp 7-9 6/0003 K1.15(2.5/2.6) l ANSWER 6.03 (1.00) b ! REFERENCE TPT SD117 'AFW', pp 10 013/000i K4.04(4.3/4.5)
#~
ANSWER 6.04 (1.00) b REFERENCE TPT Lesson P n ' Gamma Metrics Neutron Detector
- Requal Cycle IV-1985
, K6.01(2.9/3.2) l l
l l 1 l l l l I l t
, [ p aae , gugygD 8TATES NUCLEAR REGULATORY COMMISSION
, , ,g l
. o M000N il l 3
* \ 101 MAPWTTA 8TMET N.W., SulTE 2000 l ATLANTA, GEORGIA 30323 l
( 'cTGN, CONTROL, AND INSTRUMENTATION PAGE 24 j ______________________________________________________ q l ANSWERS -- TURKEY POINT 354 -86/02/03-DEAN, WH ANSWER 6.05 (1.50) ! c) False (+.5 ea) b) False . c) True
-REFERENCE -
TPT SD63 'ESFAS'r pp 32,52, FIG 14 006/000; K1.02(4.3/4.6)
.(
CNSWER 6.06 (,, J4M Bridge crane; S. _ r ' " , , : 7 ;--*- - - . '. ; ' . ; Highest obstruction in the SFP and canal'- (+.25 ea) REFERENCE (vP8W%Cf( ) TPT SD44 ' Fuel Handling System'r pp 16 NRC IE Info Notice 85-12, 11 Feb 1985 034/0003 K4.02 (2.5/3.3) ANSWER 6.07 Y
' 5M l*DD o ) ' N c ris f + ' ^a-Te s Po n s e s - W WW g b7 c)
Closes Gas Decay Tank Discharge Valve (RCV-014) Closes S/G Liquid Sample Isolation Valves (2800, 2801, 2002) f (86i W
*
- 3 Blowdown Flow Control Valves (6278 A, B & C)
Dump Valve to Discharge Canal (LCV-6265) REFERENCE TPT SD68 ' Radiation Monitoring and Protection System'a pp 34/35 000/059 & 060; EA2.05(3.6/3.9) & (3.7/4.2)
ne% UNrTSD STA788
", , (%, NUCLEAR RE00LATORY COMMISSION
.E N. !
{e j l 181 MARIETTA STREE7, N.W., SUITE 2000 ATLANTA, GEo#4NA 30323 l
/
- 6. PLANT SYSTEMS DESIGNe CONTROLe AND INSTRUMENTATION PAGE 25 CNSWERS -- TURKEY POINT 314 -86/02/03-DEAN, WM ANSWER 6.08 (2.00)
- 1) OMS Mode Control Switch in ' Low Press' (+.5 ea)
- 2) PZR PORV Control Switch in ' Auto'
- 3) PORV isolatioin valve (535 or 536) open
- 4) Power available to PORVs and PORV isolation valves .
REFERENCE TPT SD7 'RCS'e pp 49 002/0003 K4.10(4.2/4.4) ANSWER 6.09 (1.00) g (
- 1) 862 A and B (RWST to RHR) (+ M e
- 2) 864 A and B (RHST isolation)
- 3) 865 A, B and C (Accumulator Isolation)
- 4) 866 A and B (SI Hot Les Injection) r) N A c~-4 %
REFERENCE g T T gDg,1,,*4 g S'e pp 36 006/0008 K4.08(3.6/3.7) ANSWER 6.10 (1 00)
- 1) Low Gas Pressure (reagent or calibration gas) (+.25 ea)
- 2) Low Analyzer Temperature
- 3) Low Gas Flow
- 4) Analyzer Cell Failure REFERENCE TPT SD28 ' Containment Post Accident Monitoring Systems', pp 9/10 028/0008 A4.03(3.1/3.3)
ANSWER 6.11 (1.00) l c) Backleal. age from S/Gs via check valves (+.75) b) Vent the pump casing (+.4) once a shift (+.1) i I l
A% UNITS STATES
;/[> %, NUCLEAR RE:ULATORY COMANSSION AteeN u
- f tes aAAAETTA STREET.N.W., SulTE 2000 ATLANTA, oEORGIA 30323
\*ee*
- 6. PLANT SYSTEMS DESIGNe CONTROL, AND INSTRUMENTATION -5 ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, WM REFERENCE TPT EO 63, Cycle I Requal-1985 035/0108 K1 01(4.2/4.5)
ANSWER 6.12 M.avs O*d
~=5 ^'dE-a=r~troidi: - a ni m r i = + -- ~ 'A mtM- - -' -
b) Accordian type mechanism w/ a fusible leak (+.5) if hot gases pass thru ducte fusible link nelts (at approx 165 des F) releasing the damper and sealing the duct. (+.5) REFERENCE TPT Lesson Plan for Requal Cycle IV-1985, ' Appendix R Update' 086/0008 A1.04(2.7/3.3) ANSWER 6.13 (2.00) y,y 9g o) Tubine is runback at 200%/ min for 7.5 seconds M stops for.G9;T seconds then repeats cycle if cond on still exists &+-rST(+.7) b) The rod position indicating system initiates the runback (+.5) ::(1; .;
-- - - ^ - 17^% ;; :: ::d b, I;7bi e 1;t ;t:30 i^'F9155 r===> *-
=== ' ' . 5 ) ,' 4 pm;J. yf ,f 17.5FM R WMD NO(200BonoMisp7 90smirs) I"?!-y.& 4 REFERENCE TPT SD127 'Hain Turbine Control", pp 18/19 & Fig 11 045/0005 K4.12(3.3/3.6)
ANSWER 6.14 (1.00) Td
- 1) you could parallel MCCs on different units (eliminate unit separation) 4-rST h4)
- 2) the MCCs could not be in sync if paralleled causing circulating currents and create overload conditions M (-)-3)
REFERENCE TPT Lesson Plan 20-OL, APP B ' Replacement of 120 VAC Inverters', pp 5 1 Roqual TPr twG 6C(vele to -T- F-IV-1985
<TO1 062/0001 A3.04(2.7/2.9) l l
l
L '- lf% -T..cT= l
* ;y( #g NUCLEAR RCULATORY COMMl4Sl0N
'101 MAAstTTA STMET,N.W SutTE 2000
{ ATLANTA, MOAGIA 30333 .
- 6. PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUNENTATION PAGE 27 ANSWERS -- TURKEY POINT 314 -86/02/03-DEANr HH ANSWER 4.15 ( .75)
MSIV may not shut on a low steam flow situation when required (+.5) ; if there is a loss of instrument air (+.25) l REFERENCE I TPT LER of 23 July 1985 . NRC IE Notice 85-84 of 30 October 1985 035/0108 K6.01(3.2/3.6) I l l 1 l 1 l I
*g% wafseat Tes
,. 'E #g
*a NUCLEAR RE^ULATORY COMMISSION y ,
{ 101 RAAAGETTA STMET. N.W., BUITE 2000 ATLANTA, M0fMMA 30333
\*ee*
7 * "4 - NORMAL, ABNORMAle EMERGENCY AND PAGE 28
~~~~R I656E66f6AE"_66UTR6E~~~~~~~~~~~~~~~~~~~~~~~~
ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, W M ANSWER 7.01 (1.00) , O REFERENCE ~~~~~~ ~ ~ TPT E0P--FR-S ~.1 001/0108 A2 08(4 4/4.6) ANSWER 7.02 (1.00)
.d REFERENCE Wastinghouse background info for TPT E0Psr "RCP Trip / Restart'e pp 49/50 000/0748 EK3.07(4.0/4.4)
ANSWER 7.03 (1.00) d+h - REFERENCE TPT E0P-ES-0.3/0.4 000/0118 EA2.08(3.4/3.9) ANSWER 7.04 (1.50) o) False (+.5 ea) b) True c) False REFERENCE Wastinghouse User's Guide for TPT E0Ps, pp 5-12 i I
I
~ '
# % INeTE00T1738 kg NUCLEAR RE^ULATORY COMMISSION M080N N
{ 191 MAAIETTA STMET,N.W., SWITE 2000 ATLANTA,GE0404A30323 -
\..... I l
- 7. PROCEDURES - NORMAle ABNORMAle EMERGENCY AND PAGE ?? 4
------------------- ..--------------------------- 1 RADIOLOGICAL CONTROL l l
ANSWERS -- TURKEY POINT 314 -86/02/03-DEANe WH CNSWER 7.05 (1.50) (.C M 1) 2) Tavs/ Tref Deviation Alarm Overpower Rod Stop (+cB6-ea) Mg*j j 3 )- RCS - High -Delta T- - - WN) --
- 4) OP/07 Delta T Rod Stop
- 5) RCS High/ Low Tavs
- 6) Rod Bank D Low Limit Alarm REFERENCE TPT Requal Cycle II Lesson Plan-19858 NRC Generic Letter 85-05 of 1/31/85; TPT ONOP 2608.1/2; TPT SD5 " Rod Control'e pp 24 & SD7 'RCS'e pp 69 004/020; PWG-10(4.3/4.5) .
ANSWER 7.06 (1.00) (N M Centainment pressure (+.35) > 4 psis (+.15) b Containment radiation (+.35) f'IOEE5 R/Hr (+.15) REFER NCE #@ b*8T) Wostinghouse" bac ground info for TPT E0Psr ' Instrumentation Accuracy'epp 11 022/000; K3.02(3.0/3.3)% 44 Ar cop eurnM 'lP 6 % I ANSWER 7.07 (1.00)
- 1) Criticality planned within 4 hours (+.25 ea)
- 2) Boron Concentration < 300 ppm
- 3) Trip was from > 40%
- 4) Equilibrium Xenon existed prior to trip REFERENCE TPT OP-0202.2, pp 15 001/050; PNG-12(3.7/3.7) l
l see UNITED STAT 38
', ' , ' ' g NUCLEARRE ULATORYCIMMISS60N
, e REGMHl3
{ l 141 blAA;ETTA STREE7, N.W., SUITE 3000 ATLANTA,GEOAGRA 30823 l
%,e...e l 1
I l
'/ -
~nilRES - NORMALr ABNORMALe EMERGENCY AND PAGE 30
--- RasiatsaiExt casiaat------------------------ i l
ANSWERS -- TURKEY POINT 314 -86/02/03-DEANe WH ) i l ANSWER 7.08 (1.50) l
- 1) Rod can't be moved (+.5 es)
- 2) Rod misaligned by > 15 inches l
-3)- Rod -drop -time not - met -~- ---- - -- - - _ _
REFERENCE - TPT TS 3 2-3 000/0018 EK3.02(3.2/4 3) l ANSWER 7.09 (1.00) [M OM Nb An SI pump (+.25) takes a suction on Loop C (+.25) via the RHR system cnd its HXers (+.25) where its cooled by CCW (+.25) m .'&
- p t 3 (+ @
(veJ!al [la REFERENCE g g P ,0,5ga pp 5/6 , wctLad.y for keul{sN) ulcaQ,i A) 005/0008 K3.07(3.2/3.6) ANSWER 7.10 (1. 50h t amf
- 1) Unit 1/2 Cranking Diesels via Unit 3C 4KV bus (+.5 es up to 1.5)
- 2) Unit 3C Bus transformer via Unit 3C 4KV bus
- 3) 4C '
4 44 SU transformer to 3A bus O st./Yb - ww WNet - M yMe [ Nester *
~
R FE E E
# * *
- WO TPT ONOP 9408.2 t) rt.stge //t. sh WMr - c%ke4 M 84 TPT SD140 'Hain Pow r Distribution', Fig 1 - 3 e rg , g A bwG M(o F- r r /
6 /0008 A2.05(2.9/ .9) l l f
meTse st'.Tes NUCLEAR REQULATORY COMANSS60N
. . *a ,
{ Itt t4ARIETTA STAGET.98.W 00tTE 2900 ATLANTA.SECA04AStata . 7 *RnCEDURES - NORMAL, ABNORMAle EMERGENCY AND , PAGE 31 -
~~~~R I6 UE6656AE"66NTR6E~~~~~~~~I'~~~~~~~~~~~~
ANSWERS -- TURKEY POINT 384 -86/02/03-DEANe WM ANSWER 7.11 (2.50) o) -SI Annunciator DN (+.3 es response) , ,,,
-SI pumps running ,
-RHR pumps - running -- - - --- -- -- - -- 2-- , - - - --
-EDGs running .
b) -CNTMT Purge / Supply fans OFF
-Purge Valves CLOSED
-Instrument Air Bleed Valves CLOSED
-Verify Control Roon Ventilation Isolation c) -AFW steam supply MOVs OPEN
-AFW Flow Regulator Valves OPEN j REFERENCE l TPT E-0, pp 4/5 ,1 000/0073 PWG-11(4.4/4.5)
ANSWER 7.12 (1 50) . To prevent excessive depletion of RCS inventory (+.5) such that the RCP trip occurs (+.5) at a point where the break would completely uncover , the core (+.5) ! l REFERENCE Wastinghouse background info for TPT E0Psr 'RCP Trip / Restart' 000/009; EK3.23(4.2/4.3) ANSWER 7.13 a) Let t e r ing(.SN ( 2. 00 (+.5) of sub-tasks ) - b) Return to next step or sub-step on the left side (+.5) c) Honitor all remaining trees for RED terminus (+.5) and if.not encounterede suspend any ORP and perform the applicable FRP (+.5) REFERENCE Westinghouse User's Guide for TPT E0Psr pp 3-11
% ISWTM STATM T;
[ k NUCLEAR REGULATORY COMMISSION Mesoes u f tet blARIETTA STMET. 06.W., GUITE 3000 j, {- ATLANTA. GEORGIA 30838 - o
'eee*
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONSe AND LIMITATIONS PAGE 32 ANSWERS -- TURKEY POINT 384 -86/02/03-DEAN, W M i
ANSWER 8.01 (1.00) O --- REFERENCE ; TPT AP 0109.3, pp 5-6 l ~ ' ~ PNG-218 ~ 0btain/ Verify Contro1~ Procedures (3.8/4.1)
~~ ~ ~ ~~ ~ ~ ~ ~ - ~
l ANSWER 8.02 (1.00) ; b 1 REFERENCE . TPT TS 3 10-1 103/0008 PWG-5(3.1/4.1) ANSWER 8.03 (1.00) dMA . REFERENCE TPT TS 3.8; TPT IE Report 84-250-39/40
)
061/000; PNG-5(3.3/4.1) 1 ANSWER 8.04 (1.00) ) d REFERENCE l TPT TS B3 M'g.l 035/010; PWG-5(3.1/4.0) 1
[
.e,. 8,=,. +
' , ' ' , , /. .,g, NUCLEAR RE00LATORY C0000AIS840N
~~,
f 181 blAMETTA STREET,N.W., SulTE 3000 ATLANTA, GEOMNA 30333 -
'l
%..... t 3
0 ^nNTNISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 33 f e ANSWERS -- TURKEY POINT 384 -86/02/03-DEAN, W M , ANSWER 8.05 (1.50)
- c. FALSE (+.5 es) l
- b. FALSE l
- c. TRUE ]
R E F E R E N C E - ---- - - - - - - . - - - - . - - - - - - - j Cote TS, p. 3/4 0-1 and B 3/4 0-1 i TPT TS 3.0.1 and 4.0.1 PNG-58 TS Knowledge (2.9/3.9) )
l ANSWER 8.06 (1.50) c) No (+.5 es)
- b) Yes -----
c) Yes REFERENCE TPT TS 3.7.2(b)) TPT LER 85-009 of May 1985 064/0503 PWG-5(3.1/4.1) ANSWER B.07 (1.00)
- 1) Plant Supervisor-Nuclear (PS-N) (+.5 ea) l
- 2) Nuclear Watch Engineer (NWE)
REFERENCE TPT AP 0103.4, pp 9 PWG-14: Tagging / Clearance procedures (3.6/4.0) ] ANSWER 8.08 (1.00)
- 1) To verify receipt of an annunciator (+.5 ea)
- 2) To initiate corrective action in the event of an emergency REFERENCE TPT AP 0103.2, pp 3 PWG-23: Plant Staffing / Activities (2.8/3.5) l l
l l l
_d UNrfW STATES r ei t % 'S NUCLEAR REGULATORY COMMISSM)N
. . .: $$ d
.5 h rtd n l asesoN u 191 MARIETTA STREET,N.W 9UITE 2900
- ATLANTA eEORGIA30823 r
*e e* ,
d f- "'MTSTRATIVE PROCEDURES, CONDITIONS, AND LINITATIONS PAGE 34' 5
---------------------------------------------------------- C ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, W N ANSWER 8.09 (1.50)
- 1) wired speed (+.4) < 10 mph (+.1) (werger/ /6e r Jehtu)
- 2) wind direction (+.3) from the south or nor{heast (+.2) ~ '
- 3) precipitation is falling (+.5)
REFERENCE - ------ ---- - . - - - - - - - - - - - - - - - - - - - - - -- . - - - - - - - - - - TPT SD50 ' Gaseous Waste Disposal Systen'a pp 25; f 071/000; PNG-7(3.2/3.7) l l I ANSWER 8.10 (1.50)
- 1) Visual inspection of breaker position" (+.3 ea) l
- 2) Breaker light indication !
- 3) Functional Test (eg. voltmeter) l
- 4) Local (or Remote) Instrumentation l
- 5) Annunciators REFERENCE TPT AP 0103.4, pp 3 PNG-13: Conduct / Verify Valve Lineups (3.7/4.0) ,
l ANSWER 8.11 (1.50)
- 1) Classification of the Emergency (+.5 ea)
- 2) Decision to notify state / local authorities
- 3) Protective Action Recommendations REFERENCE TPT EP 20101, pp 2 .-
PNG-36: Facility E-Plan (2.9/4.7) ANSWER 8.12 (1.50)
- 1) Tavs < 578.2 des F (+.4 for parameterr +.1 for setpoint)
- 2) P:t Pressure > 2220 psigu
- 3) Rx Coolant Flow > 266,500 GPH l
l
gusteBSTATES
, neog% ,
NUCLEAR REGULATORY COMM:2SION Rees0N N i 101 ISARIETTA STREET. N.W., SUITE 2000 .l ATLANTA,et0RetA30833 . *i
..... g
[ q S. ADMINISTRATIVE PROCEDURES, CONDI'TIONSe AND LIMITATIONS PACE 35 ) ANSWERS -- TURKEY POINT 384 -86/02/03-DEAN, W H l J l REFERENCE ' TPT TS pp 3.1-7 l 002/0207 PNG-5(2.9/4 1) 3 ANSWER 8.13 (1.00) i o g9,
- 0) In a sealed envelope in the PS-N office (+.
b) The First Phone Number dialed is N office emergency telephone (+.3) and the STA will be and the activation message will be : (+.3) if the A aler is operational _g, REFERENCE TPT EP 20 , pp 10 P 4: Operate plant communications (3 1/3.3) ANSWER 8.14 (2.00) ,
- 1) To ensure evacuation of area is complete (+.5 ea for any 4) "
- 2) To rescue injured / trapped personnel
- 3) To perform operations to mitigate the effect of the energency
- 4) To determine nature / extent of the emergency
- 5) Toestablishdefinitepersonnelexclusj$bnboundaries REFERENCE TPT EP 20111, pp i PNG-36: Knowledge of E-Plan (2.9/4.7) l I
i
)
l l
. ~ _ _.
om mee UNITSD STATSS ENCLOSURE 3
/ NUCLEAR RE20LATORY COMMISSION
... . .a 9~ s l 0, 101 blARIETTA STREET. N.W,. SuffE 2000 D. ATLANTA,0E.A04A 30333 j
%o .e* ,-
1 U. S. NUCLEAR REGULATORY COMMISSION REACTOROPERATORLICENSEEXAMINATION(RFGuel) , FACILITY: TURKEY POINT 3&4 !
.(
REACTOR TYPE! PWR-WEC3 . l DATE ADMINISTERED: 86/02/03 EXAMINER: DEAN, W H APPLICANT! _________________________ INSTRUCTIONS TO APPLICANT! , Use separate paper for the answers. Write answers on one side only. ~ f Stcple question sheet on top of the answer sheets. Points for each l quastion are indicated in parentheses after the question. The passing Srsde requires at least 70% in each category and a final grade of at i Examination papers will be picked up sex'064 fer hours after loost 80%. l tho examination starts.
% OF CATEGORY % OF . APPLICANT'S CATEGORY -
VALUE TOTAL SCORE VALUE CATEGORY _.q__ _2 :: __ ___1__ ___________ ________ 1. PRINCIPLES OF HUCLEAR POWER PLANT OPERATION, THERMODYNAMICS,
- HEAT TRANSFER AND FLUID FLOW I4 b
_ 1 __ _ _1 ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
" ^ ^
_";;;;;__ ___1_00 25 INSTRUMENTS AND CONTROLS _ ___________ ________ 3. 18 0 25 0 PROCEDURES - NORMAL, ABHORMAL, ___I__0__ ___1__0 ___________ ________ 4. EMERGENCY AND RADIOLOGICAL CONTROL
' b7(r W 72.00 100.00 TOTALS FINAL GRADE _________________%
All work done on this examination is my own. I have neither given not received aid.
~~~~~~~~~~~~~~
. hPPLEC NTI5~5EGU IURE
f UNITED STATES
, /.SM'%,% NUCLEAR RE2ULATORY COMMISSION
' '. E i' ntosoN u
{o f 101 blARIETTA STREET N.W., SulTE 2900 ATLANTA, GEOAGIA 30333
*e.e*
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2 -
--- isEER55isisiEi- sEsi isisiFEE in5 FE5i5 FE5s GUESTION 1.01 (1.00)
- Which set of parameters below best describes centrifugal pump runout conditions?
- a. High discharge pressurer low flow, high power demand
- b. High discharge pressurer low flow, low power demand '__1 _
j
- c. Low discharge pressurer high flowe high power demand
- d. Low discharge pressurer high flow, low power demand ;
- e. Low discharge pressurer low flow, hiah power demand OUESTION 1.02 (1.00)
Which of the following curves (see attached page) representing Xenon concentration is correct for the given power history? OUESTION 1.03 (1.00) Which of the curves on the following page shows the e:<pected trace on the NIS Startup recorder for equal reactivity insertions in a suberitical roactor during a reactor startup? OUESTION 1.04 (2 50) The plant is operating at 30% powere turbine in AUTO (IMP IN), when loop
#1 reactor coolant pump trips. Assuming a reactor trip does not occure there is no operator action and rod control is in MANUAL, indicate whether the following parameters will be HIGHER, LOWER or the SAME at the end of the transient compared to their initial values.
- 1) #2 S/G steam pressure (0.5)
- 2) #3 RCS loop flow (0.5)
- 3) Te in loop 41 (0.5)
- 4) Th in loop #2 (0.5)
- 5) Nuclear Power (0.5)
(xxxxx CATEGORY 01 CONTINUED ON NEXT PAGE *****)
c
* , . r . *
= .
S
= ** ** s'. i.e n.. iit f
O. E' g __ - - - -. - - - - - - - SO 90 6D $6 leo $10 190 Q -- - - - -- G - W d 8- t to qq 50 80- 180 sto Ik0 C. g . 10 40 60 to too gao No a O. . to go 60 80 sto leo He TIME (Howns) t . FIGU R E
s v ed ,s'. , TInnf
~
4 ' a
# B -
a , a a a t
. t . a . .
, . . s . . n . e n . . a TIMi
=a ,
9
.c =
- c. ,.
l e
. . . . . . . . . t .
runi hNwer ( I I # # I
. . . . . . . a a e a e A a a e i t
}\O$
1 l l i l , ' - ' - - - - - - , . . _ , - , , , , , , , . , _ _ , _ _ _ _ _
'l UNITED STATES
[ # 888 gk NUCLEAR RE00LATORY COMMISSION .[ RE000N N 3 101 MARIETT A STAGET, N.W., SulTE 2000
* - ATLANTA, DECAGIA 30323
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3
--- isiiR55isiRICi! REEi isiniFER AR5 FEUi5 FE5s QUESTION 1.05 (1.50) -
For the changes listed below (treat each one independently) indicate whether the moderator temperature coefficient will become MORE NEGATIVE, LESS NEGATIVE or have NO EFFECT. (Assume all other parameters are const' ant) ' c) . Neutron flux peak _ shifts radially outward to the edge of_the core. - b) Baron concentration increases 100 ppa while core is at MOL. c) All rods in instead of all rods out. QUESTION 1.06 (1.50) An ECC is calculated for a sta'rtup following a reactor trip from 100% power equilibrium xenon (BOL). Indicate if the actual critical rod position will be HIGHER, LOWER or the SAME from the calculated position for each of the following situations. Use attached curves as appropriate and treat each case individually. o) Xenon reactivity curve;for trip from 60% is used to calculate conditions to startup 20 hours after the. trip. b) The Samarium reactivity curve is used instead of the xenon reactivity curve for startup 60 hours after trip. c) The power defect curve for 750 ppa is used instead of the 1450 ppa curve. QUESTION 1.07 (1.50) s) TRUE or FALSE: During cold plant conditions, you would expect the COLD calibrated PZR level instrument to indicate HIGHER than the HOT calibrated level instrument. (0.5) j b) Give two different conditions involving the reference leg which could result in a false high level indication on the PZR level instrument. . (1.0) l OUESTION 1.08 (1.00) What are the two primary factors that provide the driving mechanism for Natural Circulation flow? (xxxxx CATEGORY 01 CONTINUED ON NEXT PAGE *****) l
C [f ..
"j'*l fk ,dk E
g j..l: :. :'* .bi: *Tj ..
' ..{ .
{*
.-1... ;;*J .um.,.3_ ::: .
,7 ,, . ;; *{ ;. .n ' , ,
J. 4:
,;g.,..'*g. .. . .. .*; . u.o: .1:3" : ... ...7;;... ;. .*.J_ .- ..
. 7 {:!
;h';. .J. ;;.. .. . . . . ... . . ; 7 n.; 3.. :, .:. m.g m J
.. ... . 3 ,
c3 N ' ; . .
. fa s i4 .. :i: ..
. + .' .. m./ y -;r,
. - * .. * * .n::; :r- i::
.!-u- '* .. .1. :2;; . .
I . .
= -
a:-.. . er .:n- : - 4 :: H p .
- ~..... ,
- ..::. . . . ; t:. --: :: ..t, .a.
.. _ ..._ _.=
. ...:: .... ...n. _ ..
. I .
.:. u. . . . -t- ua - ".;. .:E.=:::
_ I)- .
.2. - . ... ;u:. - . - ;-;- .u- : .
.?.
.. .:n. m.: ::
-r :E: " - -:: .- ,:
.::: ;. .: u : :;-
.:- r
.:: 1:n ; h--- g 4-: :-: St g nr :..:
- -:: *:=:-: - .- -
- !".! =.. ::_;. - Lisi :!h u..i. -.i l.i.i.i i.E. P__ _- K iit ! ii". i: : E. i_'? - .-:- .- E. i. :ii.i. :E- .,
- s. Q.! . . .
r .= :. .=::n.: :.- =:.; .: :. - u : .-- :: .: :.-~u. - .jj:, ,:J.
..:. . : .CI ' u_ .... .:.: 4:: ;.;; ::::run3:j .; .:.. . .;. . :"; .-- ..
. . .._ : 7: ;t,; :,, y, , , ,,
--r-r: . ==.:: :u: :: -E:
-::: u:? ::.;r -1 :2 . 7 :: :
- - : ;, 1 af::t :: : = :ut n= . - c. :.; ;;
. == -t8 . - - _-
1
=. : .1 ' E -
.r.: =u :. .f t-a=p
- .-.. .- ...r....
J :;:: : n ;; m{ f-: -= ::...:- :: 7:1-.:..- ::: --- : p.. d
.. c
- : ._ . . .. . ._:. . :t-:.e _ _ ....
e*
!E =. 7
-] ::
4i +; =- E.a:=r ."E :i!- in * - du di:. dii J i. = -
~ .iE:-i ji: Q H.id i:E-.:: :d-EE ::_- 2. ._ j=mE j;ig :-g_
EEE @-g :.!_2_.: ji-j 4is @ ijE2i3gi; S:. .. u m _-c. _. [3j
. .~-
=
"---~~
e _., _ . . . . .__._ . . ... -_.: _..a... .... . . --_. .. . . _ ._. y. _ . ; - ~. -
=-tt: =.=- = = ....=.n..; .. . . ..- .-,~ . :. ,..__ = _..__,..
.'* _.s._u- . _ . ..c: - _==
. . . .___ -:.._.... .- =:_ ....u f
- . . .2. =- =--- .
9- i
)
*~ ~"~ '
, " . .a".*- - : :: -. . , .. . w;:4
- h % #:Es :~izz n".;. .--~:
c:;: -5: : 8 -
=u--='= - p-E:_. :: - E..*:t.:=
=='
;.::- .= .:_ _ .:. .l3.n:-= "=:- :. .
_=.
***- ~- I *- ' [
...;.~ -I=F 2. .*
2 *" * ~~ _..~."-"". Eb s _. '=. "..a __..I:- --: _n'- *.*"~. ~"J.'.' _"-l":
.= ::~.*. *. " . "_. "" ._-.3=._.: *:.
..~.___... .
*. = .
_':I"...- =..." .!._.P.*.-*
- . 7. *::...0.
.s _
_ . " .g # y g__. . -.._ _g _ . . ._ . _ ..=- ._-
... _ . . .- - --- - m = RE == u-: = - - -:i ~_ 3 %
e ==_===..u--- _:- :- ._. _ . _ _ _= ..<*-* _..-
- I
.ll=_ '.. 2_ -. . _ __ " "__.~ .' y:I.:..
-= - _ ._
~d... *.".;. .:.i. . .. :: -*. *. _*
.. _ . . . __. __ . _.. i O _._.* ..** ___.'32 : .". _.:.7..-.:*._.._'O. .- .- .* . . . ' ~_.. . _ .~ ~ _ - -" . _ . . .~.
g _
.. ~ . . . .
= -._. .-.... . ... . -
*t;;.". . _ - _-r y
.P_- '.
.,: :_= :_:: . . = ... . -.-- " "_.. . .
_0- '_"I .. .:.-= .=. 2 ... --...*.= - ...
~ -.
-.i _. .
^
f
~1 **=_'-."l'-"
-- " - '- -- ~* " - --a . ;: ,= r_ = ~:~=; . .._
_ . . ._ - - ';=" ~ "~*^-n.
== : -- .J::.;; -. . ;. :=;.:
Q', _ " : . .._g_g=~'- g
. .~=
'= 'u" _
~ -
Q g
.._* = _=_
;.p..g
?- . . . .=___.:,.: _ .* __;=-".:-- -" : --"
__ ' ~~-~'
,. ,,,a. .
3
- :~_
..# f.;__ == :=~ **** "
.--;- - - - ' . ._ := ._
~ * - * ':-- _. _ _
. 3 .g ,
~.;, ._,
- .g. _
. i.m.__.
- - _ _c
. e
-**- - _ PU j
/
OP _ E m - W .W W l a .am ._
, , w f
e r,,,, 4 ,, e
.e
. l.._
A. E
= - -
D .- - M m W e i er g 1..
~ ~ .
w:-; I
.aw ...mm.
1 -
.A e -in._.m h 6
.m
-, p g
E M
.4 . i to - :-
4 -
-- n.
h _
. g --.
W o $ W
-# ;m mm u
um _-,E g M_ -- S PM
.m 6 C -:
" gm i y -
M
^
* -- - p" . _ . . -;.,_ ,--;;
__.j,
. . --
- IL- :J-
--w --* r - : .; .:1::. .e_. . ;;;
m _- .- .
- _-- y , ,
M , , V .. L . 2 - I A. e m e n w co -
POWER DEECT Yh % REACTOR POWER
. LNT3 CYt2.E X 90L *
~
16 750 ppm l i ! 1100 ppm l 14 1450 ppm D 1650 ppm l i 11 i 10 M . 9 3 1> 6 5 l 1 1 3 . l l 2 - 1 . s Reactor Power (R15:1) G
'Ni]
POWER DEFECT VS, % REACTOR POWER
. LNT3 CYCLEX SQL .
16 750 ppm 15 1100 ppm 1450 ppm 114 13 12 11 1650 ppm
$- 10 M .
.]
1> 6 5 4 3 2 - 1 . E. Reactor Power (R15:1) , _. h
, . *2 ?'.~
1- ,
- i 1
N WORD 4 V5. M 94JTDOWN-ptX. { g . l ( .
;)
j; E l-soo ; l-l
/ l \~
Joo
- ]
s . 400 t o, 20 40 60 80 loo 120 14o 160 -180 200 220 24o 260 too
. Time After Shut &rwn Hours
~
.h
.=-
. 4@w.:;
3 .. .
1 g_..._...._...._...._.......,_....... _ _
.....,.y . .;
_ g . _ .. . .yt . ,. .-3
; .. % _ a
. ')& w.
. .. . 4
. . ~. . . . . . .
., ,o, . ,. .. ...m . .,. .. ,. ,
.t- .
,MM.m. ... ... .... m
- 4. - t. 1.. . : ,, . , i,,
m.
"4 ., . . ., .. _ _ -
. .... . . . ... .. . t .. . L, ... m u,- . ,
.i
, , i,,
- n. 4 , . . .
. , ... . is ,
.... . .1. .t 1
4 .
..1.. .
g.t, b....
.. .. . . . . ., .t .... . 31 .. , .
..... .... . .n.
,. .. ... g, . g, . L.
. 1,:,1., ,....
g-g,
. . v1..I" .
.._-'.t g I ,.. ...
,U . ._ . . . .... . . . . . . , ., ,, ---
., 1
..a t.t!.+ .*'.t...,...
.. 2
-r-
.*2.. J.2 1,. .. .. --
.... ... ... .... ... .. . .. .... .... .. .. .... .13 . ... .. . , i 1 X . .n.
- . .... . 1 . - .
.. . . _ . . . . .. .. . .. *t:
.rn. 3L .
..12 .~
J .
-.. .:2
. $r,. . .
, . . +. ..... . . . ,_ .. ..
n,. . ..
. s' . i .. - -
- c. .- -
. ..t . .- . ..2 . - .- :*43 .. .
m .*!'.. g (. p:.. . 33: - - -
. ,. . . 1. . .. - -
. .3 . ...,... . . . 2-
. . . .M 1._. . . -
333,. . .;: .-
-t 221
. 2.
w
. 4S.,
. .c ...
_.4 I .n .- . . . .. . . - _ l........_. . .. ... ai d *
, , qg i
. .- .M W f.
;9 --
=* t.*y .,*e F4. a tv ^ '.;_ " --- "
l *a -y .+.= _ 0, ,, ,
. --M m w a g N3 - u lIe g N " ^
E "" . _ " E 4. a w _ M_ m . M M m . v K M . ..
^
W .
. - -M M E .-
- - -.J .
] E M .
a M . 1.E . . . 2 1M . m M 'E -
. - - .J .
_X .
-~rw.
R R RJ E _A _ . , .m . ( _. ._. a - N . .
-. .. _n.
.m 1 J nm IM i.
.. Qg W
~'
i E - 3 N*8 w. ., - m . m, . M
-M .
; - M
- -m -
'M _ W
.--.-A
.-_.m,- MMI
, . - M
. . - ,M .
m V - AM '
'M"g m
M . h R_ M M-h
. M -MM . M m
W - M! 1 E I . . EE IAM IA. m
- M I w '.WM mm w
I.'m
MM '
W - - r M g h M EM M i1 . F u 1 M _ a M M . l g M3 I M . _M h M a r M'FE I1LMJ' m I I m I w 1 I a y , a . I JM i M _ 'E a Lj.' 1 M g- -E M _ M 1K 1 .
.M
-d' e
' M,Ma- .
W II M K- E m
.%Md h
M M-m-m I r M m M MM . M M m I I I 1 IM.. m M _ AM M A iMM W - M _MIMM M-m m I :K__~~ iA . I_.r L Im.F: I- a M 1hFsf.P 1III h . A ILX k . ' k I ; ML~ ' 2 -W I I E-E I
'n_ yiXAMM
- m KmI 1 wm . nwyAMm
. I I
y u I .m . - m -
.F I .
.-- -J F _ a m E_m I
I w- - y I I 1m . h - . h r . m I I i r g l.ME II I J m I 3 'M m .j IE ,( 1 I 1I I L
. IM M I E a.
/ I 6
I I ._I_ II I
-s ,. .
M m m = X1 N.i _ l X I I 'M""'hMk I
'IL
' :M I1 I . L u
m a II ,_ a h
- h _h y
.f l
IIT Ts .1 an MEN , I rm -,g a g, wMLM-j, y T 1I l . . I
~'
L - I N'.- :K I II i 1 I , g f
._,+..f I
-I , %;tE:u m -.-++- -- 4 I :
( M 3 3 S X R h 2 E c * *
. R cxnso m a
h N = e, .. r.
... ..,i, .[ . .,4 4 ., se ' ,.q.g. .... . . . , . . p e. . ..
. , ., tg""
q u _... -Cf ,s. .
- s. ..
.. ... f.."
.o. 4.
i.
- i. T' t,a..
~ . . a. .. . - -. . . . .. . . . .-.,~a . a ,! ,i w ,<
.i.,
2
.** .1: .. it'. .
. ..u n
^
~3,t. . ..
.%. i 1:, . .,.i
.._. .. ~.. .-.-a
.. . . 2:. . . .
.. ,. M ,. !. .. ... . 1. . .
,It 1 1 .o. 1.2.. 1. ... ,
m . .,
- t. .....
... ..., ...o...
. ....) . .... . ... . . . .
. 1. .
. 1..
2***. _. n . .
.r
. a.2*J .l.. .
o.,, 2
,. .=.=l
,p .
. . . . .9. . , .
. - , . .7.L.
. 4. 4 .. . . . .
... . , .. t. . g.,. .. , . , . ,
l.. ...... e .
. . ... .. . _. . u ,
. "3 1 ..,.. 7"_* .
r ., :q' ,==3 s ._ .. - - .
*T
- I* .t" . ... . . . ,
. .. .- _. .. . . . ".U .- -
= .-a, .
- m. .22*. . . .,
. ...- 22. .
... _ .,,, ., 4..... . . . .. . . ...
1
.Iti .
,1m. a,, ,.. < .,. ,,
.-a a.... . .- -1 o . . .1 . . ... . , . ..2.*.. . ..-. . .. a *t..- .a. . .. .
2.fr ;f: .. .. .... aa.,,..,, X...=--' -.- , *.*
.. , , .,.o .. -
6.,,. 4...
. ... 2. .. .,.
.i
.1:'" n 311 9'
p** ,,.%
^ #
3 pg - ..6. . ._ ; -.
,,,,,,g
- 4. 1.. . a-. . '. . . 6
=2r . .4.',< ,.4.
*::*' .1 o...
.a. ... .....
.a.
.a66
- a. .
.L4
. a. .,.
..~. ...
,.1 . ..,, ., . , ... ..5., -. . ,. ,
*J
.q
...i,. .
m
,... .. ..., _. .,4 _. _.
-- . . . . , . ,..- 6+. <
- < .i g
[, .".
~ ED.1 .
-- :i - .
A g- ,
. .. ..< . u ; ,.
'-- .~
/'
.. ___ _- _ ,,m"
. ., . +. .
.R
-. . ..m. _ -_ .- y . .
- I c.e -_
9 I
.w_
Y I & 4,+ I!.
. I1 13 .
Os " sg , s 1E R E. R. A i I M . I I " 1 IY II I I . I I I I I I L g I I I
!. g g k , H A__ m b
IaI I I I I I
- I 1 Y I I I I k 3 - . I
~
I 1 I w'E - T I I 1 g 1 1 1 g I I I 1 1 1 y y 1 s' I ( B I I - II II I I 1 , ,
. m 3 , [
Suu l, 1 .
%ir , l
, 1 , I
_m
,1 z
- 't u
.I I
m -t 1 1 1 , z 1- m _J 'l~ 1 1 1 1 I 2 1 1 1 1 1 1 1 2@.- r(, .,
- h. +t- "-+1*
1 1 1 1 1 r 1 1 1 1 - ,_ 1 y ,
; ( sh - ,
m 1 , , w - Qn '1 111 , 11 1 l1 1 1 1 !
, 1 MM.q gg' Ur--- Ih. .
- . : _ _ . . _! F $ t+: - x
?-
- d Ptt+
- r .-*.. ;** t =_::nO _. _:.. : - ._-
J:::"!:ti N ..- ..
- W?ttr.M-"n:::tt;:n :e+v ;tt+"+<mt+-
. ;;;; ;-- :: - : 1 1 -
_W1-.mm..y FN.-. M4 ++ - - -- --
% .oe.m ; 4ttt"; u , ;
1
; 1
.. m y .
m.
-. +m 1-1 1-- 1 1 1 i i 1 : i - ~ .. . . ++ d +. M4P .
-- b b . _ S _____ ___ __
#8%g UINTED STATES
', O k n' NUCLEAR RE ULATORY COMMISSION M000N N
{ l 191 MAR 4ETTA STMET,N.W., tulTE 2000 ATLANTA, GEORGIA 30333 - i,t s/'
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 4
-~~~infiR557RARICi? Risi fiiniffi AR5 FEUi5 FE6Q QUESTION 1 09 (2.00)
Unit 3 has just restarted following a refueling outage while Unit 4 is naar EOL. Answer the following regarding the differences in plant response ' batween the two units (explain your answers)! , _o)_ At_a steady. power _ level _of_10EE(-8) _ amps _during a_startupe equal __ __ __ reactivity additions are made (approximately 100 pen). Which Unit will have the higher steady state startup rate? bl At 50% powere a control rod (100 pen) drops. Assuming NO RUNBACK or OPERATOR ACTION, which Unit will have the lower steady state Tavs? QUESTION 1.10 (1.25) There are two effects that cause differential boron worth to change over core life. List these two effects, their relative impact on differential baron worth and indicate which effect is the overriding factor. QUESTION 1 11 (1.50) o) What is the definition of Shutdown Margin (SDM)? (1.0) b) If a stuck rod exists while the reactor is at powere what adjustment, I if any, must be made to the SDM calculation? ('0.5) 1 1 QUESTION 1.12 (1.00) Operating Procedure 0202.2, Unit Startup, states that during a reactor startup, "a non-uniform increase in count rate will occur' when withdrawal I of Shutdown Bank A is commenced. What is the reason for this phenomenon end what is the approximate change in count rate that is observed? GUESTTON y ? ' 1,251 ' h UNIT 4 is just critical in the intermediate range when rod D-4 (which was at 140 inches) begins to withdraw at 32 steps per minute. Assuming a differential rod worth of 5 pcm/ inch, what is the SUR 60 seconds into this rod withdrawal accident? Show your calculations. (***** END OF CATEGORY 01 mraxx) l l
f nee UNIT 10 STATES f, NUCLEAR RE20LATORY COMMISSION l
.'.')- %*A ntoeoM m l
.f 101 MARIETTA 8TREET, N.W., SUITE 2000 l
ATLANT A,0EoAGIA 30323
'5% , . . . . . sN
)
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 5 QUESTION 2.01 (1.00)
Which of the following statements correctly describes the RHR System lineup then HOT LEG RECIRCULATION is established? (assume both trains of RHR are available)
- a. Both. trains are used to supply hot les recirculation exclusive'ly.
- b. One train is used to supply hot les recirculation and the other
' train is^used to~ continue cold les recirculation.
' ~~ ~ ~ ~ ' ~
- c. Both trains supply both hot and cold les recire simultaneously.
- d. One train is used for hot les recire and the other train is put in standby (ie. recirculates from RHR HXer outlet to RHR pump inlet).
QUESTION 2.02 (1.00) - Which of the statements below regarding Unit 3 AFW pump steam supply valves en a loss of the 3A 4KV bus voltage is correct? _,
- a. Steam supply valves from all 3 S/Gs open.
- b. Steam supply valves from A and B S/Gs open. ._
- c. Steam supply valves from B and C S/Gs open.
- d. Steam supply valvcs from A and C S/Gs open. -
- e. No steam supply valves will_open as it takes loss of voltage on both 4KV bus 3A and 3B to cause the valves to open.
i 1 (max *x CATEGORY O2 CONTINUED ON NEXT PAGE *****)
U g' *% UNITED STATES NUCLEAR RECULATORY COMMISSION gc 3,
.. o namoNu t01 MARIETTA STREET,N.W.,006TE 3000
{e l ATLANTA, GEoRetA 30333
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 6 QUESTION 2.03 (1.00) -
Wh of the following describes the NORMAL BACKUP s rce of water to the Fire ection System if electrical power was lost to the Fire Protection ' System pump
- a. The Back Service Water pump takes a suction on the Raw Water' Tanks and is ' ed up to the Fire Protection pump suction. ,
b.- A spoolpiece is ins ed between the Fire Protection System and the Service Water Pump arge.
- c. The Screen Wash Pump Discharge connected to a nearby fire hydrant using a hose. -
- d. The Elevated Storage Tank is lined up to pr de gravity feed to the Fire Protection System by opening valve 794 really closed).
QUESTION 2.04 (1.00) Which of the following flowpaths describing how power is normally cupplied to a typical vital instrument bus is correct?
- a. 480 VAC from vital bus, rectified to 125 VDC, inverted to 120 VAC, and supplied to instrument bus.
- b. 480 VAC from vital bus, transformed to 120 VAC, and supplied to instrument bus. -
- c. 125 VDC from battery, supplied to battery bus, inverted to 120 VAC, and supplied to instrument bus.
- d. 480 VAC from vital bus, rectified to 120 VDC, and supplied to instrument bus.
(xxx*x CATEGORY 02 CONTINUED ON NEXT PAGE xxxxx)
pnee% UNITED STATBS
,, k NUCLEAR RESULATORY COMMISSION A8040N N l 1
101 MARISTTA STAttT.N.W.,808T8 3900
.l- ATLANTA,GE0AelA 30333 ,
e
*e..*
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 7
_______________________________________________________ l l QUESTION 2.05 (1.00) l 1 Answer the following questions regarding the instrument air system TRUE or FALSES o) Control valves that cross-connect the UNIT 3 and 4 Instrument Air compressors will automatically supply instrument air to the other i unit if that unit's air compressor failse as long as the supplying l -- - - - - unit 's air - pressure remains above psis. --- --- l 1 b) If BOTH UNIT 3 and 4 lose instrument aire the service air from UNIT ' 3 and 4 is the preferred backup source of air over the instrument ~ air from UNITS 1 and 2. QUESTION 2.06 (2.50) . Match the RCS penetrations in Column A with the appropriate RCS loop cassent listed in Column B. (column B items may be used more than once but only one response per penetration) Column A Column B '
- a. Excess Letdown 1) Loop A cold les'
- b. Pzr Surge Line 2) Loop A hot les
- c. Alternate Chstging 3) Loop A intermediate les
- d. PZR Spray Line 4) Loop B intermediate les
- e. RHR Suction 5) Loop B hot les
- 6) Loop C cold les
- 7) Loop C hot les GUESTION 2.07 (1.00)
Fill in the blanks in the statement below regarding the Standby Steam Generator Feedwater Pumps (SSGFP): Besides being used during startup and shutdown, the SSGFPs are also used as a(n) __________. These pumps are located adjacent to the __________ tank and take a suction on the __________. tank. Flow to the S/Gs is controlled by the _________ valve (s). (xxxxx CATEGORY 02 CONTINUED ON NEXT PAGE xxrms)
UtelTED STT.TES ! # "*% h, k NUCLEAR RE'ULATORY COMMISSION ma,0,, 0 tl'"l (e $ 101 MAAIETTA STMET.N.W SutTE 2000
- ATLANTA,OE0 Atla 30333 Y
/ -
?)
g *eee* / o
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8 I
QUESTION 2 08 (1.25) l
- Answer the following questions regarding CVCS limitations and precautions! h a) The Oxygen concentration in the VCT cover gas must be maintained below f
_____%. (.25) [ 9 b) During any change in RCS Baron concentratione at least-one of what'two ______ major systems / components _must be_in operation? (1 0) - 9 GUESTION 2.09 (1.00) What is the purpose of the following precautions associated with operation of the Reactor Coolant Pumps? o) Do not open 41 Seal Leakoff isolation valves until RCS pressure is greater than 100 psig. (0.5) ; b) Do not open 41 Seal Bypass valves until #1 Seal Leakoff valves are open with > 50 psid across 41 seal. (0.5) j
-QUESTION 2.10 (1.50)
The sodium tetraborate decahydrate added during the injection phase ofter a LOCA will eventually be distributed by the Containment Spray j System and raise the Containment Sump pH to 8.5. What are the 2 reasons for establishing this elevated pH in the containment?
% C'JEC TIC N 2.11 (2.GG) - dk %
List 4 of the 5 Design bases for the ECCS Cooling Performance following a - LOCA as stated in 10CFR50.46. QUESTION 2.12 (1.00) What are the 2 purposes of the interlock that prevents the LTDN isolation valves from opening or shutting unless all three orifice isolation valves are shut? (xxxxx CATEGORY 02 CONTINUED ON NEXT PAGE xxxxx)
L. UNiTto STATES $
/ p mee %", NUCLEAR RE2VLATORY COMMISSION I
- - Aro.oN . F-
.! ~
5 1St MARISTTA STREET.N.W., SUITE 2000 *
- ATLANTA, GEOA04A 30323 -
*tr
*ene*
a
, C
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGF 7 ,
QUESTION 2.13 (1.25) - o) Preventing steam binding of AFW pumps at Turkey Point is a major concern. What is the potential cause of this steam binding? (.75) b) If AFW pump casing temperature is 150 des Fr what action is.frequired to reduce the temperature? (Include the frequency of the action) ('0.5) , UESTION 2 14 Il "^b[bo) ' What indication does a control room operator have that a fire damper has actuated? (0.5) b) Explain in detail the design features which allow a fire damper to auto close when required. (1.0) (xxxxx END OF CATEGORY 02 xxxxx)
/pha UNITED 5TATES }
'e,, NUCLEAR RECULATORY COMMISSION 5 ,y' '
AEpon u { 191 RAARIETTA STREET, N.W., SulTE 2000 ATLANTA.GEOAGIA M333
*1r
*e. *
- 3. INSTRUMENTS AND CONTROLS PAGE 10 _
QUESTION 3.01 (1.00) Which of the following correctly describes the actions of the AFW Flow 1 Centro 11ers AFTER receiving an initiation signal? 1
- a. Since the AFW flow control console Hand Indicating Controllers ~ .
(HICs) are set to a predetermined flow rate, the control valve ~ will RAPIDLY OPIN to this predetermined setting.
- b. Even though there is a pre-set flow rate from the HIC positioners, the flow control valve is initially driven to the FULL OPEN '
position due to the large error signal between the HIC and the initial 0 GPM actual flow measured.
~
- c. Due to a long reset time in the flow control circuitry, the large -
~
difference between the HIC setpoint and the O GPM actual flow ' measured initially, the valve will SLOWLY OPEN to the setpoint. , ,
- d. Due to a large flow measurement ' spike" above the pre-set position on the HIC, the flow control valve will initially STAY SHUT, then 0 as flow stabilizese OPEN. SLOWLY to the pre-set. position. -
TUCOTIC" M2 ( 1 : ^ o)- Which statement below correctly describes operation of the GAMMA-METRICS Nautron Flux Monitor in gamma flux fields between 10,000 and 1,000,000 R/hr (ie. high radiation fields).
- a. The monitor is not designed to operate in such high level radiation fields.
- b. The monitor will operate satisfactorily in these radiation levels, but an adjustment should be made to discriminate against the higher samma flux.
- c. The monitor will operate satisfactorily, but the output signal from the detector will not increase linearly due to lack of voltage saturation in the detector.
- d. The monitor is designed to operate as well in this high a level samma flux as it does in much lower radiation fields.
1 (xxxxx CATEGORY 03 CONTINUED ON NEXT PAGE xxxxx) l I l
ase UNiltoSTATBS NUCLEAR RE20LATORY COMMISSION <
** REGION N
. {[ .
101 MAAIETTA STREET. N.W., SUITE 2900 ATLANTA, GEOAGIA 30333 -
\*.* 4 I___ _______ ____________
QUESTION 3.03 (1.50) ; Indicate whether there are in 2 or 3 SELECTABLE detector inputs for each of 1 the following parameters utilized by the S/G Water Level Control System. ~ c) S/G Level c b) Feed Flow M c) Steam Pressure . - QUESTION 3.04 (1.00) l Indicate whether the following situations would cause the steam dump system , to ARM ONLYr ARM & ACTUATE or HAVE NO EFFECT! o) PT-447 (1st stage impulse pressure for load reject signal) faiIs LOWe Mode Control in Tave moder Tref > Tave by 6 des F. - b) Turbine Trips, Mode Control in Tavs moder Loop A Tavs fails HIGH, the Steam Dump Control Reset switch is HELD in the ' Bypass' position ~ , CUESTION 3.05 (1.50) Answer the following questions regarding ESFAS-TRUE or FALSE! o) In order to generate a "P' signal, 2 out of 3 Hi Containment Pressure OR 2 out of 3 Hi-Hi Containment Pressure signals are required. - b) The S/G Delta P 'S' signal actuates when 2 out of 3 S/G pressure detectors.for 1 S/G are 100 Psi GREATER than 2 out of 3 Steam Line Pressure detectors. c) A Reactor Trip in coincidence with a Low Tavs will result in Main Feed Water control valve closurer but the Main Feed Water pumps will still be running if they were operating at the time of the reactor trip. (xxxxx CATEGORY 03 CONTINUED ON NEXT PAGE xxxxx)
4l UNITED eTATSS
/.pn met % NUCLEAR REQULATORY COMMISSION .
[' RE KHd M 101 MARIETTA 8TREET, N.W., SulTE 2000 [
- i ATLANTA 0E0A04A 30333 -
\p.....
l l
- 3. INSTRUMENTS AND CONTROLS PAGE 12 !
y QUESTION 3.06 (2.00) M0tch the interlock. descriptions in Column A with the appropriate logic rcquired to cause rod withdrawal to be blocked in Column B. (column B ; items may be used more than once) l COLUMN A . COLUMN B ~
,;l a)-Power - Range - High- Flux e-103% - power -----
~
-- - - - - - 1. - 1/ 2 ~ ,
- 2. 2/2 b) Overtemperature Delta T rod stop 3. 1/3 - ,
- 4. 2/3 c) Intermediate Range High Flux ~ 5. 1/4 ~
- 6. 2/4 d) Power Range Rod Drop 7 3/4 1
QUESTION 3 07 (1.00) l
- l With.the pressurizer level control selector switch in position III/II, h
l en instrument failure causes the following plant events in sequence (Assume no operator actions taken): ~ l
- 1. Charging flow reduces to minimum 1
- 2. Pressurizer level decreases
- 3. Letdown secures and heaters deenergize~ )
~
- 4. Level increases until high level trip Which instrument failed (II or III) and in what direction did it fail?
QUESTION 3.08 (1.50) a) Give the location and the number of UV relays that must be energized I to initiate bus stripping on a Loss of Off-Site Power. (0.5) b) To ensure needed vital equipment starts on a Loss of Off-Site Power following an SI that has been RESET, the operator can perform what two actions? (1.0) OUESTION 3.09 (1.20) When nuclear power has been increased above the setpoint for permissive P-10, the operator can manually block three protective features. List the THREE features that can be blocked. (xxxxx CATEGORY 03 CONTINUED ON NEXT PAGE xxxxx)
g% UNITED STATES h*
# %* NUCLEAR RECUtATORY COMMISSION
8 , nE0eoM u
{ t et uAn:ETTA STREET, N.W., SufTE 3000 ATLANTA, GEORGIA 30333 1 1
\*eee*
__I"'IsMENTSANDCONTROLS PAGE 13
]
OUESTION 3.10 (1.80) List the 6 reactor trips,whiAh are enabled / blocked by the reactor trip cystem interlock P-7. QUESTION 3.11 (2 00) What are the 4 conditions which must be met for the overpressure nitigation - cystem (OMS) stitus 1TshTs to WONT~~
' ~~
q j QUESTION 3.12 (1 50) The Bus clearing relays set-up the permissive to use the startup .s transformer from unit 3 to supply the 4A 4KV Bus. Before the cross-tie breaker (3AA22) can be shut, what 5 conditions are required to be met? 1 1 QUESTION 3.13 (1.00) The Alternate Source Transfer Switches associated with the recently 1 installed 120 VAC inverters have key locks to prevent 2 switches of the j ocae channel being selected to ALTERNATE at the same time. What are the l the purposes behind this administrative key contrel? 4 l l (xxxxx END OF CATEGORY 03 xxxxx) l l l
unrfteST'.Tes 2
/
- r. NUCLEAR RESULATORY COMMISSION
'I
',[ 'g nateon n .
g tt1 MAhlETTA STREET,N.W.,9ttfTE 3000 g - ATLANTA,080 Atla 30333 _ ___ b _ __ _I__ __ I_$ _ ___I_ __ _ ___ b ~~ RADIOLOGICAL CONTROL q
. i QUESTION 4.01 (1.00) b The Response Not Obtained for the first immediate action of E0P-FR-S.1 ' ' Response to Nuclear Power Generation / ATHS' is to manually trip the f reactor. If the reactor will not tripe thent
- a. Place rods in Manual and insert them int's the core.
~
Trip the turblirielhd verify steam dumps open.
] '
b.
- c. Energency borate the RCS.
- d. Dispatch operator to locally trip reactor.
QUESTION 4 02 - (1.00) Which of the following would cause the greatest biological damage to a man? g
- a. 0.1 Rad of Fast Neutron. .
- b. 1 Ren of Canna.
- c. 10 Ren of Beta.
- d. 0.05 Rad of Alpha.
QUESTION 4.03 (1 50) . Answer the folowing questions regarding E0P usage TRUE or FALSE: a) If a Function Restoration Procedure (FRP) is entered due to an ORANGE Critical Safety Function (CSF) condition, and a HIGHER priority ORANGE condition is encountered, the original FRP must be completed prior to proceeding to the newly identified FRP. b) Unless specified, a task need not be fully completed before proceeding to a subsequent step as long as that task is progressing satisfactorily c) If a procedure transition occurs, any tasks still in progress from the procedure which was in effect need not be completed. (xxxxx CATEGORY 04 CONTINUED ON NEXT PAGE xxxxx) l
% WINTS ST!.TES ('
NUCLEAR REZULATORY COMMISSION
- ,'. 8 \*"
f RE00001I
,{ '
101 MARIETTA STREET. N.W., SutTE 2000 ATLANTA,OE0AGLA 36323 -
\*e.e* l l
l l
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 15
~~~~
RA5i5E55f5AEc5NTR6E~~~~~~]~$~~~~~~~~~~~~~~~ QUESTION 4.04 (2.00) < 1 Fer the following paragraphe choose the correct words from the options * - I given after the paragraph that correctly complete each blank. - When paralleling a diesel generator to the gride the generator ___'__ voltage should be a the line voltage. The diesel l generator is synchronized to the grid by observing the .
- I synchro pointer as it moves slowly in the ___b___ direction 1 and closing the generator breaker when the pointer is ___c_'__ )
the vertical position. The power (MW) output of the generator l is then_ raised by adjusting the ___d___. q Choose from the following! -
- s. Iower than / equal to / higher than (0.5)
- b. slow / fast (0.5)
- c. 5 minutes to / at / 5 minutes after (0.5)
- d. governor control / voltage regulator / stator cooling (0.5) 'f QUESTION 4.05 (1.00)
Fill in the blanks in the statements below regarding startup and normal pcwer operations: a) After leveling reactor power at _____ amps to take critical rod height datar power is increased to 2% with a steady state startup rate of no more than _____ dpm. b) Control banks should be manipulated to adjust Tavs to within _____ des F of Tref before shifting rod control from manual to automatic. c) The reactor shall not be made critical with a difference of greater than _____ pcm between the projected critical height and the ECC rod position. QUESTION 4.06 (1.50) ONOP 1008.2r ' Excessive RCS Leakage', identifies numerous methods by which RCS leakage may be determined. List the 5 radiation monitor alarms which could be symptoms of RCS leakage. (***xx CATEGORY 04 CONTINUED ON NEXT PAGE *****)
tNeTW STATES NUCLEAR REOULATORY COMMISSION I
.'.*! nemona f
,{ tot muuusTTA sTasst m.w., surfs sees ATLANTA,GE0nelA 30823
')
,l N
%,..... j i.
- 4. PROCEDURES - NORMAL, ABNORMALe EMERGENCY AND PAGE 14
--- E A5i5E55iEAE E5 EVE 5E------------------------
QUESTION 4.07 (1.00) - 4 List the two conditions (including setpoints) which determine Adverse ) Ccntainment conditions. - l QUESTION 4.08 (1.50) ; What are the Unit 3 RCO's immediate actions if the word is passed " Fire in the control roome shift personnel report to assigned control room evacus- 4 tion stations' ? --
~
QUESTION 4.09 (1.50)' E0P-ECA-0.0e " Loss of All AC Power' has the operators check if the RCS is isolated as one of the immediate actions. How is this step accomplished? QUESTION 4.10 (1 50) - After' Natural Circulation has been established, what 3 indications are conitored to determine RCS COOLDOWNr according to ES-0.2e " Natural . Circulation Cooldown'? l QUESTION 4.11 (3.00) . List ALL the immediate action sub-steps from E-0, " Reactor Trip or Safety Injection
- that allow you to accomplish the following immediate actions: _.l c) Check if SI actuated (1.2) b) Containment Ventilation Isolation (1.2) c) Verify AFH pumps running (0.6)
GUESTION 4.12 (1.50) - During a small break LOCA (SBLOCA), it is required to trip the RCP if the trip criteria are met. If forced flow through the core pr omotes cooling, why are the RCPs tripped. (xxxxx END OF CATEGORY 04 xxxxx) (xxxxxxxxxxxxx END OF EXAMINATION xxxxxxxxxxxxxar) l l
- c. . m mw
. evt)/(Energy ta) 2 l* ,
o o og s e V,t
- 1/2 at I
E = ac Kt = 1/2 av t
, , gy, , y g)fg g ,gy g , g ,-At
,t =.egn vf = V, + at * = 6/t a = an2/t1/2 = 0.693/t1/2 y , ,3p A= aD 2 t 1/2'#I
- EII D'/III)) b 4 [(tjjg)+(t,))
at = 931 an -
. m = V*,A, -Ex g ,= a' th I*I*o Q = aCpat .
6 = UAa T I = I ge'"* rwr = v fah I = I,10**
- L TYL = 1.3/s P~=~P*10'"#I*I HVL = 70.693/s
, = , e s/T ;., . ;
SUR = 26.06/T SCR=5/(1-Kyff) SUR = 26 lp + p)/(lfeff-p) - CR, a 5/ ( 1 - K,ff,)- - --- SUR = 26e/s* + (s - e)T s CRj (1 K ,ff)) = CR 2 II
- Ieff2)
T = (s*/s) + [(a - sy Ie] - M = 1/(1 - K,ff) = CR /CR j , T = a/(e - s) M = (1 - K ,ff,)/(1 - K ,ffj) T = (s - e)/(Te) SOM = ( - K ,ff)/K,ff a = (K,ff-1)/K,ff = AK,ff/K,ff s* = 10 seconds . T = 0.1 seconds ~I - e = [(1*/(T K,ff)] + [s,ff /(1 + IT)] ! Idjj=1d
, = (seV)/(3 x 1010) Id 2 ,2 gd 2 jj 22 E = sN 2 R/hr = (0.5 CE)/d (meters)
, R/hr = 6 CE/d 2 gf,,g)
Water Parameters Miscellarieous Conversions 1 gal. = 8.345 lem. I curie = 3.7 x 1010dps 1 ga:. = 3.78 liters 1 kg = 2.21 lbm 1 ft< = 7.48 gal. 1 hp = 2.54 x 103 Stu/hr . Density = 62.4 lbg/ft3 1 m = 3.41 x 100 atu/hr Density = 1 ge/ce' lin = 2.54 cm Heat of vaporization = 970 Btu /lem *F = 9/5'C + 32 Heat of fusion = 144 8tu/lbm ,
*C = 5/9 ('F-32)
, 1 Atm = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-lbf 1 ft. H 2
= 0.4335 lbf/in.
ec 2.718 W" '
- - ,w g "
__m...._ _ _ _ , , , _ , , _ . _ - . . . . - - - _ _ , _ _ _ . _ , - . - , - -
i
** Wehame.C'/lb IstheW. Sh@ Empepy, temp a F
% y wow top suem tow tap einem wenn av , ges y
% %* % h % & N % %
at s.0s859 a01602 3305 3305 4.02 lens 30764 0A000 2.1sn 2.1873 32 SA1602 3948 2948 8.00 10733 10763 DA061 2.1706 2.1767 35 SS 0.09991 0.01602 2446 2446 S.03 1071A 1079A 0A162 2.1432 2.1504 40 40 0.12163 0.0262 2.1164 2.1426 45 45 0.14744 0.01402 2017.7 2037A 1344 1068.1 1081.2 0A1602 1704A 1704A 18D5 1065.3 1083.4 0.0381 2A901 2.1262 to 90 0.17796 0.0655 2A091 2.0946 80 80 0.2561 0.01603 12074 1907A 28.06 1059.7 1087.7 03629 0.01605 868.3 368.4 38D5 1054.0 1092.1 0A745 1.9900 2.0645 70 90 0D932 1.9426 2D359 90 00 0.5068 0.01607 633.3 633.3 4844 1044.4 1096.4 0.01610 468.1 468.1 58.02 1042.7 1100A 0.1115 13970 2.0006 to 90 0.6981 0.1295 12530' 1.9825 Boo 300 03492 0.01613 350.4 350.4 68.00 1037.1 1105.1 265.4 77.98 1031A 1109J 0.1472 1A105 1.9577 110 110 12750 0A1617 265A OA1620 20325 303.26 8737 1035.6 1113 4 0.1646 1.7603 13339 330 ISO 1A927 15112 130 0A1625 157.32 157.33 97.96 1019 2 11172 0.1817 1.7295 13e 22230 1A010 1A005 340 0.01629 122.98 123A0 107.95 1014.0 1122A 0.1985 140 2AD92 150 3.718 OA1634 97A5 97A7 117.95 1008.2 1126.1 0.2150 - 1A636 1.9606 350 OA1640 7727 77J9 12736 10022 11302 02313 1A174 1A487 360 180 4.741 62A4 62.06 137.97 996 2 1134.2 0.2473 1.5422 1A296 176 OA1645 1,,70, _ . 5393 g ___90M - SOR - -148.00 - 990.2 - 1138.2- 0.2631 - 1.5400 - 1 2111 San-_ 9.340 0A1657 40.94 40.96 15844 9P4.1 1142.1 02787 12148 1.7934 330 190 300 11.526 0.01664 3342 33.64 16829 977.9 1146D 0.2940 1A824 1.7764 300 210 210 14.123 0.01671 27A0 27A2 178.15 9714 1149.7 0.3091 1A500 1.7600 0A1672 26.78 26A0 100.17 970J 1150.5 OJ121 1A447 1.7568 212 212 14.696 OA1678 23.13 23.15 188.23 965.2 1153.4 0J241 1A201._1.7442 220 220 17.186 230 0.01685 19.364 19.381 198.33 958.7 1157.1 0.3388 1.3002 1.7290 230 20.779 24.968 041693 16.304 16.321 <
- 208.45 952.1 1160.6 0.3533 1.3609 1.7142 340 240 29A25 0.01701 13302 13A19 215.59 945A 1164.0 0.3677 1.3323 1.7000 250 250 260 35A27 0e01709 11.745 11.762 228.76 938.6 1167A OJS19 1.3043 1A462 360 4L356 0.0171: 10.042 10.060 a3a.95 931.7 1170.6 02960 1.276e 14729 270 270 300 300 49.200 0A1726 S.627 8.644 '249.17 924.6 1173A DA098 1.2501 1.6509 390 57A50 OA1736 7A43 7A60 250A 917A 11762 OA236 1.2238 1A473 290 300 67.005 0.01745 6.448 6.466 269.7 910.0 1179.7 OA372 1.1979 1.6351 300 310 77A7 OA1755 6.609 6.626 280.0 9025 1182.5 OA506 1.1726 1A232 310 320 39A4 0.01766 4A96 4.914 290.4 394 2 1185.2 0.4640 1.1477 1.6116 320 340 117.99 0.01787 3.770 3.788 311.3 878.8 1190.1 0.4902 1.0990 1.5092 340 153A1 OA1811 2.939 2.957 332.3 362.1 1194.4 0.5161 1A517 1.5678 360 360 CA1836 2.317 2.335 353.6 8445 1198.0 04416 1A057 1.5473 300 3a0 195.73 400 247.26 0.01864 1.8444 1A630 375.1 325.9 1201A 05667 0.9607 1.5274 400 305.78 041894 1.4808 1A997 396.9 306.2 1203.1 0.5915 03165 1.5080 420 420 0A1926 1.1976 1.2169 419.0 785A 1204.4 0.6161 03729 1A890 440 440 381.54 460 460 466.9 04196 0.9746 0.9942 441 5 763.2 1204.8 0.6405 03299 1A704 480 566.2 0.0200 0.7972 02172 4645 739.6 1204.1 0.6648 0.7871 1.4516 480 680.9 0.0204 0.6545 0.6749 487.9 714.3 1202.2 0.6890 0.7443 1.4333 500 500 0.0209 0.5386 0.55 % 512.0 687.0 1199.0 0.7133 0.7013 1A146 520 520 812.5 0.0215 0.4437 04651 536.8 657.5 1194.3 0.7378 0.6577 1.3954 540 540 962.8 0.0221 0.3651 0.3871 562A 625.3 1187.7 0.7625 0.6132 1.3757 560 SED 1133.4 0.0228 0.2994 0.3222 589.1 589.9 1179.0 0.7876 0.5673 1.3550 580 580 1326.2 0.0236 0.2438 0.2675 617.1 550.6 1167.7 0.8134 0.5196 1.3330 600 600 1543.2 620 0.0247 0.1962 0.2208 646.9 506.3 1153.2 0.8403 0.46S9 1.3032 620 1786.9 640 640 2059 9 0.0260 0.1543 0.1802 679.1 454.6 1133.7 0.8666 0.4134 1.2821 0 0277 0.1166 0.1443 714.9 392.1 1107.0 0.8995 0.3502 1.2498 660 660 2365.7 0.0304 0.0808 0.1112 758.5 310.1 1069.5 0.9365 0.2720 1.2086 GoG 6M 2708.6 0 0366 0.0386 0.0752 822.4' 172.7 995.2 0.9501 0.1490 1.1390 700 700 3094.3 0.0508 0 0.0508 906.0 0 906.0 1.0612 0 1.0612 705.5 705.5 3208.2 TABLE A.2 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (TEMPERATURE)
A.3
l Weisme.9tapn gew,y, gespe gas,,,y, gi je , y g g ,,go jin
.,I.-
.89g8 y tretee Eme, Sneem Water Esop Steam Weler Evap Steam C'aerr tesem 5-e, s 0, a, %, s 4, % 5
- v acen 32.01s 0.01602 3302A 330eA e.00 len& son & e a.nsn e.lsn e nor1A mase6 ;
e.se n.0e3 oment e9455 s946A Sm 19732 toms oA061 a.1705 a.3 m 6 sAs teet.3 ale ' e.15 4!.453 0 01602 2004.7 2004 7 13.50 1967.9 BeslA 0.0271 2.1140 2.1411 13.50 3025.7 e.15 ; E20 53.160 0 01603 15M.3 3526.3 2122 30635 1084.7 0 0422 SD7 M 2.1160 2122 3028 3 s20 SJO 64.484 0.01604 3039.7 3039.7 32.54 80b7.1 3089.7 0.0641 2A168 2A000 32A4 3032A &a0 ' SA0 72269 0A1606 7924 792.1 40.92 3052A 1093.3 0D799 1A762 2.0H2 4032 1034.7 OAO l a.5 79.586 0.01607 641.5 641 5 47A2 304B6 1996.3 0.0925 1.9446 RA370 47A2 1036.9 &5 S.6 85.218 0 01609 540A N0.1 5325 1045.5 1098.7 0.1028 1.9186 2Atl5 5324 1038.7 SA i A **se. e* t? s - 90d09 < 0:01619e 466 93. d6644M -4410 40427. 18003 . :0.3 A e'. I 20 H p S A M B, As< locr1040.Sc at : . L 03 94.38 OA1611 411.67 411A9 62.39 3040.3' 3102.6 0.1117 1A775 1A970 62.59 3041.7 SA S.9 98.24 OA1612 368.41 368.43 96.24 3038.1 1104J 0.1 M 4 14606 13870 9624 1042A E9 1A 101.74 0.01614 333.59 333.60 89.73 1036.1 11052 0.13M 13455 1A781 80 J 3 1944.1 SA RA IMA7 OA1623 173.74 173.M 94.03 3022.1 31162 0.1780 1.7450 13300 94AS 30513 3A SA 141.47 0.01630 113.71 118.73 109.42 3013.2 11224 0.2009 1A854 13064 109A1 1056.7 3A 4A 15236 OA16M 9053 9044 120A2 2006A In2FJ 0.2199 1A428 13626 12020 30602 43 E.0 16224 OA1641 73.515 73.53 130.20 1000.9 3131.1 02349 1A004 13443 130.38 3063.1 SA 4A 170.05 0 21645 61.967 61.98 138.03 9962 11M.2 02474 1.5820 latte latal 1065A' SA 7A 17634 OA1649 534M 53A6 14423 992.1 1136 9 0.2551 1.5587 1A168 14431 1067A 7A
. SA _ 182 86 _ 0.01653 47.328 47.35 15037 988.5 1139.3 0.2676 1.5384 13000 15034 10892 ED 42A0 - 15630-- 985.1 ~1141 A 0.2760 15234 1.7964 15628 10703 SA 18827 0.01656 -~42.385 94 -
Se 19321 OA1659 3SA04 38 42 36126 982.1 1143.3 0.2836 1.5043 1.7879 161.25 1072A 34 14.696 212.00- 0A1672 26382 MA0 180.17 9703 1150.5 0.3121 1A447 1.7568 100.12 1077A 84AB6 l 15 213.03 OA1673 26.274 26.29 181.21 960.7 1150.9 0.3137 IA415 1.7552 181.16 1077A 15 20 227.96 0.01643 20.070 20A87 196.27 960.1 1156.3 0.3358 1.3962 1.7320 196.21 1082.0 20 30 250.34 OA1701 13.7266 13.744 218.9 945.2 1164.1 0.3682 1.3313 1.8995 218A 1087A 30 40 26725 0 01715 10.4794 30.497 236.1 # 933.6 1169A 0.3921 '1.2844 1.6765 236 3 1002.1 40 50 281.02 OA1727 8.4967 5.514 250.2 923.9 1174.1 0.4112 1.2474 J4586 250.1 1095.3 50 00 292.71 OA1738 7.1562 7.174 2622 915.4 1177A DA273 12167 1A440 262 3 100BA GD 70 302.93 CA1744 6.1475 6.205 272.7 907A 11804 0.4411 1.1905 1A316 272.5 1100.2 70 i 30 312A4 CA1757 5.4536 - 5A71 232.1
- 900.9 1183.1 0.4534 1.1675 1A208 M1A 1102.1 30 90 320.28 cal?66 4A777 4A95 290J 994.6 1185.3 OA643 1.1470 1A113 290A 1103.7 90 300 327A2 OA1774 4.4133 4A31 298.5 308.6 1187.2 0A743 1.1384 IA027 298.2 11052 300 120 341.27 0.01789 3.7097 3.728 3124 877A 3193.4 0.4919 1.0960 1.5879 312.2 1107.6 320 140 353 04 0.01803 3.2010 3.219 325A 364.0 1193.0 0.5071 1.0681 1.5752 324A 1109.6 340 360 363 55 0.01815 22155 2334 336.1 859 0 1195.1 0.5206 1.0435 1.5641 335A 1111.2 He 180 373 08 0 01827 2.5129 2.531 346.2 8503 1196.9 0.5328 1.0215 1.5543 345A 1112.5 380 300 331 A0 0.01839 2.2689 2.287 355.5 8402 1198.3 0.5438 1A016 1A454 3543 1113.7 300 250 40097 0.01865 1.3245 13432 376.1 825A 1201.1 0.5679 0.9585 1.5264 375 3 11153 350 300 417 3h 0 01889 1.5233 1.5427 394.0 808.9 1202.9 0.5882 0.81223 1.5105 392.9 1117.2 300 350 411.73 0.01913 1.3064 13255 4092 794 2 1204.0 0.6055 0 8909 1A968 408.6 11181 350 400 444 60 0.0193 1.14162 1.1610 424.2 780 4 1204.6 0 6217 08630 1.4847 422.7 111E 7 400 450 456.28 0.0195 1.01224 12318 437.3 M 7.5 1204A 0.6360 CA378 1A738 435.7 1118.9 450 500 467.01 0 0193 0 90787 0 9276 449.5 755.1 1204.7 0.6490 03148 1A639 447.7 1118A 500 550 47094 0 0199 0 82183 0.8418 460.9 743.3 1204.3 0.6611 07936 1A547 456.9 1118.6 550 600 48E20 0 0201 0.74962 0.7698 471.7 732.0 1203 7 0.6723 0.7738 1.4461 469.5 1116.2 600 703 .503 08 0.0205 0.63505 0.6556 491.6 710.2 1201A 0.6928 0.7377 1A304 488.9 1116.9 700 80", 51821 0 0209 0.54809 0 5690 509.8 689.6 1199 4 0.7111 0.7051 1A163 506 7 1115.2 300 900 Ei! .95 0 0212 047963 0.5009 526 7 669 7 1196 4 07279 0.6753 1.4032 5232 1113.0 900 1000 5
- 4.5B 0.0216 0.42435 0 4460 542.6 f 50 4 1192.9 0.7434 06476 1.3910 $30 6 1110.4 1000 1100 SLE 2d 0.0720 0.37863 0 4006 557.5 631.5 1189.1 07578 06216 1.3794 553.1 1107.5 3100 0 0223 0 34013 0.3625 571.9 613.0 11848 0.7714 0$969 1.3633 566 9 1104.3 1200 3200 1300 l 667.19 E77.42 0 0227 030722 0.3299 585.6 594.6 II80 2 0.7843 05733 1.3577 5801 1100 9 1300 1400 $$7 07 0 0231 027871 03018 598 8 576 5 1175.3 0.7966 05507 1.3474 592.9 1097.1 1400 3500 59620 0.0235 02b372 0.2772 611.7 550 4 11701 0.8035 0 '.283 1.3373 605.2 1093.1 1500 2000 635 80 0.02*.,7 016766 0.1883 672.1 466.2 113B.3 0 86.~'t 04256 1.7881 662.6 10GS 6 2000 2500 65d 11 0 02r4 0 10209 0 1307 731 7 361.6 1093 3 09139 0 3206 12345 118.5 1032.9 2500 3000 695 33 0 0343 0 050/3 0 0850 8018 218.4 10203 0 9728 0 1291 1.1619 782A 973.1 3000 32'>8.2 70147 0 0508 0 0 050d 906 0 0 906 0 1.0612 0 10612 875.9 875.9 37082 TABLE A.3 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (PRESSURE)
A.4
Deenpoestuse,F
., ,At>speese.
#*' 000 900 000 900 000 000 8000 into lago 1800 8400 3000 100 300 000
( e SA161 302A 462.3 SilD 571.5 631.1 000.7 3 8 88 00 3N07 llM.7 SMla 3395 6 18361 1984 5 001.74 s 0.1395 34609 3.1362 2.1722 3J237 8270s 3J144 e 0.0161 78 14 90.N 902.N 33421 IM M 3380s $8001 all.M 37336 305.75 St?JD 3g343 32833 359.45 6 6 48 01 3146.6 liteA 3241J IPet 2 1335.9 13s4.3 1433 6 1843 7 36M.7 Int 6 7 36396 34033 174SA 3g0 3
'l (14224) s 0.1795 1874 LAM 9 13943 3.0460 SADM t.1969 R.1776 2 2159 22621 RJ0M SJIM 33509 SJSil 3A301 St.03 67A4 63A3 et 00 N 90 ton 06 91 97 87 98 24 let30 180J6 196.72 e 0.01M SS34 44.93
, 30 a 68 02 11146 6 11937 IP40 6 12s73 3335 5 1954 0 1433 4 1483 5 llM 6 3506 6 36396 9003J SMFA M) (19.' 2 1) s 0129511.D28 14593 1.9173 13692 3.Olu 2A603 f.1011 2.13M 2.1757 22101 2.2430 32744 3J046 3J33 I e 00161 0A160 79 999 33.M3 37.985 41.986 45.978 49 964 S3.946 57.9M 61.905 65382 09458 73333 77A07 35 6 48 04 148 09 1192 5 1339.9 3387J 1335.2 1383A 1433.2 1483 4 15344 15865 5639.4 1893.2 INFA 80034 913AS) e 0.1995 02940 lAIM 1A720 13242 13717 2A!M RA663 2AD46 3.1309 f.8653 3.1982 3 3297 32509 3J3 o 93141 SL6166 32356 35A38 38.457 31A66 34A65 37AIS 40.447 43435 46430 40A05 Sim 86379 ( 88.05 18811 3191 4 1239.2 1286D 1334.9 53A3 5 34323 5483J 1834J SIMJ 36303 333.1 SMPA Im3 30 4 O27.96) s 0.1995 02940 IJ006 1A397 3A821 13397 BAS 36 BA344 RA628 3A091 3.13M 3.3666 3.3979 32382 3572 e 0.0161 0A166 11.036 IPAN la.las 13.485 37.195 13299 30199 31A87 33.lM seest 3Engs 37A75 m.le ' de 6 48.10 16a ll 11M 6 ItMA 1285A 13334 1382 4 3432.1 1482.5 1533.7 15853 1638A 1992.7 SM7A 100BA 067JS) s 0.1295 0.2940 1A992 1.7608 13143 1A624 IA065 134M 13060 SA224 2A569 3A099 3.3334 3.1516 3.1007 e OA161 0.0166 7297 -- 8.354 - 9A00 -le425 31 A38 12A46 13A50 14A52_ ISA52~ S&450..l?A48_ ISA46_39A4 00 6 G8.15 16820 1181 6 1233.5 12832 13323 1381.5 34313 34813 1533.2 35853 1638.4 legt.4 1747.1 3002A (392.71) s 0.1295 0.2939 14492 8.7134 1.7651 13168 IA612 13024 1.M10 13774 2 Alto RA450 sap 66 3.3005 3.1350 e OAl61 0A166 0.0175 6218 7Als 7.794 S.560 9 319 10.075 30329 11.981 12331 last 13229 34477 i 30 & GS.21 168.24 269.74 1230.5 1281.3 1330.9 1300 5 1430.5 1481.1 1532A 1984 3 36380 letta 17468 Nota (312.04) s 0.1295 0.2939 0.4371 14790 1.7349' l.7842 13209 12702 130B9 1.M54 1.9000 3A131 3A446 3A750 3.8041 e 0.0161 0.0166 0.0175 4 935 S.5A8 6.216 6333 7A43 S OSO S.455' 9298 9A00 lea 60 3 500 llA80 See h 68.26 168.29 369 77 1227A 1279.3 1329.6 51379.5 3429.7 1480 4 1532A 1584.4 1637A 3314 17463 302.2 (327A2) s 0.1295 0.2939 0.4371 1A516 1.7038 3.7506 13036 13451 12839 13205 13552 13883 34199 3A802 SA794 e 0.0161 0.0166 0.0175 4.0786 4.6341 S.1437 S.8831 6.192S 6.7006 7J000 7J006 12119 8.7130 92134 9.7130 330 A es31 108.33 36931 1224.1 1277A 1328.1 1378 4 1428.8 1479 3 ISSIA 3583 3 3637.1 3512 17462 Nota (34127) s 0.1295 0.2939 0.4371 1A286 1AS72 1.7376 1J829 33246 IA635 13001 12349 13000 13996 3500 SAB02 e 0.0161 0 0166 0.0175 3.4661 3.95M 4.4139 43585 SJ995 S.7M4 6.1709 EGOM 7AS49 7A052 7m46 S3233 340 & GS.37 168 38 269 55 12208 1275.3 1326A 1377A 3428 0 8479.1 15303 1983 4 36367 ISOS 1746A 3001J (353 04) s 01295 0.2939 0.4370 1.6085 1A686 1.71M 1J652 13071 13461 1382S ISIM 13508 BAB25 3Al29 3A421 e 04161 0 0166 0 0175 3.0060 3 4413 33480 42420 44295 5A132 SAD 45 S.7741 E1522 64293 '53086 7JS11 300 6 68 42 364 42 26939 1217A 1273.3 1325.4 1376 4 1427.2 1478.4 3530.3 1982.9 1636.3 1880.5 17464 IS01A -! (p63.55) s 0.1294 0.2938 0 4370 1.5906 14522 1.7039 1J499 1J919 13310 1A678 13027 13359 43676 1.9000 2A273 e 0 016! 0 0166 0 0174 2 6474 3 0433 3A093 3.M21 4.1084 4.4505 43907 S.1299 5.4657 5A014 E13E3 SA704 180 e 68 47 168 47 269 92 1213 8 1271.2 1324.0 1375.2 1426.3 1477.7 1529 7 1582.4 1635 3 1890.2 17453 1801.2 (373.081 s C.1294 0.2938 0 4370 13743 14376 IA900 1 7362 1.7784 1 A176 1.8545 12894 1.9227 13545 1.9849 2A142 e 0 0161 0 0166 0 0174 23598 23247 3A583 3.3783 36615 4.0008 4.3077 4A128 4.9165 S2191 S.5200 52219 900 4 68.52 168 51 269 96 1210 1 1269.0 1322E 1374.3 14253 1477.0 15291 1581 3 1635.4 1889 2 1745A 30003 f (35130) s 0 1294 0 293B 0 4359 1.5593 1A242 1.677G 1.7239 1.M63 13057 IA426 13776 13109 13427 13732 2A025 e 06161 0 0165 0 0174 00186 2.1504 2.4662 2.6872 2.9410 3.1909 3 4382 36837 3 9278 4.1709 4.4131 4.6546 250 h 68 66 168 63 27C 05 315.10 1263 5 13190 1371.6 1423 4 1475 3 1527.6 1580 6 1634 4 1688 9 17442 1800.2 (400 97) s 01294 0.2937 0 4365 03%7 1.5951 1.6502 1.6976 1.7405 1.7801 1.8173 1A524 13d58 13177 13482 13776 I e 0 0161 0 01E 5 0 3174 0 0186 1.7665 2.0044 2.2263 2A407 24509 2 8585 3 0643 3 2688 3 4721 3.6746 3 2764 300 A 68 79 1 % 74 27u14 375.15 1257.7 1315 2 1368 9 1421.3 1873 6 1526.2 1579 4 1633 3 3488 0 1743 4 1799.6 (417.35) s 01294 0 2937 0 4337 05%5 1.5703 1A274 14758 1J192 13591 13964 1A317 1A652 13972 1.9278 1D572 e 0016! 0 0165 0 0174 0 0186 14913 3.7028 IE973 2 0332 2 2652 2 4445 24219 2.79E0 2 3730 3.1471 3.32C5 68 92 163 85 270 24 37521 1251 5 1311.4 13662 14192 1471 8 1524 7 15782 1632J 1687.1 1742.6 1798.9 i 350 a (431.731 a 01293 0 2936 0 43G7 0 56fA 1.5483 1.6077 1.6571 13009 1.7411 1.7787 1.8141 13477 13795 1.9105 13400 e 00161 0 0166 0 0174 0 0162 12841 1.4763 1.6493 1.8151 1.9759 2.1339 2.2901 2.4450 2.5987 2 3515 2.9037 400 a 69 05 16897 270 33 375 27 12451 1337.4 1363 4 1417.0 1470 1 1523 3 1576 9 1631.2 16S6 2 1741 9 1793 2 (444.60) 51293 02935 0 43 % 05%3 1.5782 1.5931, 1.6406 16850 1 7255 13632 IJ968 1A325 18647 1.8955 8.9250 e 0 0161 0 0166 0 0174 0 0186 0 9919 1.1584 13037 1.4397 15708 1 6012 1.8256 1.9507 2.0746 2.1977 2.3200 500 6 69 32 1 % 34 270 51 375 38 12312 1299 1 1357.7 1412 7 1466 6 1520 3 1574 4 16291 1684 4 1740 3 1796 9 (457.01) s 01292 02934 04304 0 5t00 1 4971 1550S IM23 1 65/8 16993 1 7371 13730 IA069 13393 1 A702 12908 ( TABLE A.4 PROPERTIES OF SUPERHEATED STEAM AND COMPRESSED WATER (TEMPERATURE AND PRESSURE) A.5
Abe press. 1**f*' store,F W
.
- C/eg la.
.. pl. temp) 300 300 000 000 600 400 700 000 000 8000 3500 3300 3300 3400 3000 f . e Obic! 0 01% 00174 0 0lu 0 7944 0 94 % 107M 3.3892 3.300s 14093 3.5160 1AFil 3.72b2 3A2s4 3.9300 000 6 09.b8 340 42 270 70 376 49 121b t 1290 3 ISSIA 3408.3 3M3 0 lblFA lb71.9 3627A 36824 37383 3796.5 4886.20p s 0.1292 SJ933 04M2 OMb7 34600 33329 3M44 343bt 347H 3.78b5 3.7517 3.78b9 BA184 3A494 3A732 e 0.0161 t el u 0.0174 00186 SAPO 4 02928 0 9072 3A102 3.1078 12023 32944 IJ8bt 3A7b7 1.5647 3A530 300 6 6984 169.65 270 89 3461 487.93 178t h IM56 3403.7 lebte 3bl4 4 1494 4743 3660 7 3737.2 3794.3 g5032)s 01291 DJ932 0emo 0SMS 06489 1.5090 3M73 34154 3Ab80 34970 3 73M 3.7679 3A035 3 8318 3A617 e OA161 0 0166 0 0174 00186 0 0704 0 6774 0 7823 0349 OM31 3.0470 1.3789 32003 IJ885 1.3669 1A446 000 6 70.11 36930 27127 3 4 73 48738 1273.1 1939.2 3399.1 14b5 8 3511 4 lbH t 1622 1 3678 9 3735 0 3792.9 5 18.2.) s 0.1290 02930 0.4h8 OMb2 0488b IA869 15484 33980 34413 34807 1.714 3.M22 3.7851 13164 33464
~ e OA161 CA1H 00174 0 0186 0 0204 0 5869 0 6858 0.7783 0 8504 0.9262 0 9998 1A720 1.1430 12131 3.2825 000 6 7037 170.10 271.26 3 4 34 48733 1260 6 1332.7 1394.4 1452.2 lbC4 5 3664 4 3620 6 3677.1 1734.3 3791.6 5 31.95) s 0.1290 0.2929 0.4357 0.5649 0.6481 1A6b9 1A311 3.5822 IA263 I E 62 3.7033 1.7382 3.7713 1.8028 33329 ~ e OA161 0.0166 OAl?4 0.01M 0.0204 0.5137 06080 0.684 0.7603 03295 0A9% 0.M22 3.Oru 3.0001 3.3529 3000 6 70A3 370.33 271A4 375A6 487.79 1249J 1825A 33894 ~3444.5 3504A lb61.9 3614 4 14 5.3 17323 37903 gbe4.h8) s 0.1380 0.2928 OA%b 0.M47 OA876 1A40 3.5149 3.5677 3A326 34530 1A805 3.72M 3.7589 1.7905 1A207 e 0A161 0.0166 0A174 OA185 0.0203 04531 0 5440 0.6108 0.6865 k?SDS 03121 03723 0.9313 0 9094 3A468 1100 & 70.90 170.M 271 A3 376.08 487.M 1237.3 1318A 3384 7 3444.7 1502.4 1559.4 1616 3 16733 IF31A 3789.0 SM23) s 0.1269 0.2927 OA353 0.5644 04872 3A259 1A996 3.5542 34000 34410 3A787 1J143 1.74M 3.7793 3A097 . _ _ . - - - e 03161 SA166 CD174 0.0185.0.0203 0.4016 0 4905. 03615 . 04250 0A845 0 7418 02974. 03519 - 0A055 0.9584 1308 8 71.16 170.78 27122 376.20 447.72 1224.2 1311 5 1379.7 1440 9 3449 4 35M 9 1614.2 86714 17294 1787A (567.19) s 0.1288 02926 OA351 0.5642 CA MS 1.4061 1A851 3.5415 34883 14298 3E79 12035 IJ371 1.7693 3.7996
, 0A)f,1 0.0166 0.0174 0.0185 0.0203 0.3176 0.4059 OA712 03282 03809 0.6311 0 6798 02272 0.7737 0A195 1400 & 71A8 171.24 272.19 376 44 487.65 1194.1 12961 1369J 1433 2 14932 1551A 3609.9 1668 0 3726.3 3785A S87.07) s 0.1287 02923 0.4348 0.5636 04859 1.3652 1A575 1.5182 15670 34096 3.6484 1AS45 3.7145 1.M08 IJ815 e 0.0161 0.0166 0.0173 0.0185 0.0202 0.0236 0.3415 0.4032 0 4555 0.5031 05482 R5915 0.6336 0.6748 0.7153 1600 6 72.21 171.69 272 37 376 69 487.60 616.77 4279.4 33585 34252 1486.9 1546.6 3605 6 1464.3 17232 1782J 910437) s 0.1296 02921 0.4 M 4 0.M31 0.6851 03129 1.4312 1A963 1.5478 13916 14312 1A678 3.7022 1.7344 1.7657 e 0.0160 0A165 00173 OA185 0.0202 0235 0 2906 0.3500 0.3988 0.4426 0 4836 03229 0.5609 0 3900 0 6?43 3800 a 72.73 172.15 272.95 376.93 487A6 615.58 3361.1 15472 3417.1 1480.6 M41.1 1601.2 34602 1720.1 3779J 5 21.'32) s &1284 0.2918 CA341 0.M26 E68*3 0A109 SA054 1A768 1.5302 1.5753 1 AIM 3A628 1.6876 1.7204 1.7516 e 00160 SA165 OA173 0.0184 0.0201 0233 0.2488 0.3072 03534 0.3942 0.4320 0.4680 0.5027 RS365 0.5695 2000 6 7326 172.60 273 32 377.19 48733 614 48 1240.9 1353 4 3408.7 3447.1 1536 2 1996.9 3657.0 1717A 1777.1 0635 30) s 0.1283 0.2916 0.4337 0.M23 04834 0A091 1J794 1A578 13138 1.5603 3.6014 14391 14743 1J075 1.7309 e 0.0160 DA165 CA173 E01s4 0.0200 cm30 0.1681 R2293 02712 0.3068 0.3390 0.9692 0.3000 0.4259 OA529 2500 8 7457 373.74 274.27 377A2 487.50 612.08 1176.7 1303.4 33862 3457.9 1522.9 1585.9 1647A 37092 3770A 0668.11) s 0.1280 0.2910 0.4329 0.5609 0 4815 0.8048 IJ076 1Al29 1A766 1.5269 3.5703 1A094 1A456 1 4796 1.7116 e 00160 CA165 0.0172 00183 0.0200 0 2 28 0 0D82 0 1759 0.2161 0.2484 02770 03033 0.3282 0.3522 E3753 3000 a 7553 17t. 88 27522 378 47 487.52 610.08 3060 5 1267.0 33632 1440.2 1502A 1574A 16355 3701.4 IM1A GE9f.33) s 0.1277 0J904 0A320 0.5597 0.6796 0 8009 3.1966 3.3692 1A429 1.4976 1.5434 1.5641 1421A 1.0561 1.6888 e 0.0160 0 0165 0.0172 0.0183 RO199 0.0227 0.0335 0.1548 0.1987 0.2301 0 2576 02827 03065 0.3291 0.3510 3200 & 76.4 1753 275 6 378.7 487.5 609 4 8003 1250.9 1353 4 1433.1 1503A 35703 3634A 1698.3 3761.2 (105.08) s 0 1276 02902 0.4317 05592 0.6768 0.7994 0.9708 3.3515 1 A300 1 A866 1.5335 1.5749 14126 1.6477 1A806 e 0 0160 0 0164 0 0172 0.01E3 0 0199 OM25 0.0307 0.1364 0 1764 0.2066 0.2326 0.2 % 3 0.2784 0 2995 03198 3500 & 77.2 1760 276.2 3791 487.6 608 4 779 4 1224 6 1338.2 1422.2 1495 5 1%33 16292 3693 6 17b72 s 0.1278 0.2899 0 4312 0 5585 0.6777 02973 0 9508 1.3242 1A112 1.4709 15194 1.5618 1.6002 14353 1.6691 e 0 0159 0.0lfA 0.0172 0.0182 0.0198 0 0223 0.0287 0.1052 0.1461 0.1752 0.1994 02210 0.2411 0.2601 0 2783 4000 4 76.5 17.2 277.1 379.8 487.7 606 5 763 0 3174.3 1311.6 1403 0 1481.3 1552.2 1619.8 16852 1750.6 s 01271 02f 93 0.4304 0.5573 0.6760 0 7940 0.9M3 1.2754 13807 3.4461 1.4976 15417 1.5812 3.6177 1.6516 e 0 0159 04164 0 0171 0 0181 0.0196 0 0219 0.0268 0.0591 0 1038 0 1312 0 1529 0 1718 OIS*0 0 2050 0 2203 5000 a 81 1 7.79.5 279 1 3812 488.1 604 6 7460 1042 9 1252.9 13(4 6 1852 1 15291 16009 1670 0 1727.4 s C.1*M 5 f.2861 04287 0.5550 0 6726 02880 0 9153 1.1593 1.3207 1 A001 1A582 1.5061 1.fA81 1.5863 14216 e 0 01'A 0.0163 0 0170 C0160 0 0195 0 0216 0.0256 0 0397 0.0757 0.1020 0.1221 0.1391 0 1544 01684 0.1817 60C0 & E3 7 181.7 281.0 382 7 AP8 6 602 9 7361 9451 1168 8 1323 6 1422.3 1505 9 1562 0 16542 17242 s 0 1258 0 2670 0 4271 0 5528 06693 0 7826 ,0 9026 1.0176 12615 135?4 1.4229 1.4743 1.5194 1.5593 1596.2 e 00158 0.0163 0 0170 0 0180 0 0193 0.0?!3 0 0248 0.0334 0 0573 0 031 A 01004 01160 0129E 01424 0.1542 7000 6 EC2 184 4 283 0 384 2 489 3 601 7 729 3 901.8 1124.9 1281 7 139? ? 1482.6 1%31 1639 6 1711 1 s 01252 0 2859 0 4256 0 5'.07 0 6G63 0 7/77 0 8926 1 0350 12055 131>1 11934 1.44v6 1 4938 1.53'S 1.5735
'I U LE A 4 PROPERTIES OF SUPERHEATED STEAM AND COMPRESSED WATER (TEMPEPATURE AND PRESSURE) (CONTINUED)
A.0
70 . __ ,. ,., ,, o, .. .. M ' ,, ,. .. 4 m ,, i I O F U) I? -;
, ,, n , , .. . . -
Alk,t i M NH N ._ j yi n n, -- gi m -
,,,, i lF) ID ri k l' N1 /I' W in r.
171 wi i , ui i, .
,, M (M k %n I N // f%J _
im r1 41 n ii n r z e lIl m J i
,,, rn i Mz n- f// % J _-
na - zw
. ,,,, drM @
/YE, 0 $,
h i R/ x i sn N///W , c - - - s ,,,, .i
,,,, r N ,WM t -t i IA F/ M D,,,,!
1 JMA TR& y;(;n# M 7 t "'s 1_ ,,,, ppf ;//O R w 'z y. i I
.. If6H E h%WM/4 '"
_ MUN YA & 7 W/ ' ~ _ DM WhW9 4 f, C '
. MMPO Y>dD 5M D '
UbyY M 5< M
~
_ l
. MM& M ' ~
l
.. WMENDY ~
. 8WMMDV 10 a.: 12 a.3 34 a.s s.y 14 a .'
88 a. a.2 af l l FIGURE A.5 MOLLIER ENTHALPY-ENTROPY DIAGRAM l A.7 l
. -0 h
a v .. ., y PROPENTIES OF WATER Donohye :
~
M*Nti PSIA Temp Saturated 3400 2500' 3000 3000 2100 3200 3300 PF) Liquid _1000 S2Att St.900 St.93 82.951 88456 82 82.414 82.837 82A46 82A67 80 42.38 82A5 82.75 82.774 St.798 Stat 2 82446 8247 6238 ( a2.#e e2A27 et.44s e2406 82Ase 100 s1. sos m185 st.371 s2.Se0 80.549 00.588 00A87 00.006 00.102 300 90.118 80.314 80.511 80.53 67A13 57A38 57.850 57AR2 57A08 - 300 67.310 57A37 67.767 57.79 54.342 M.373 64A29
~400 63.951 E903 ---54A18 54.249_M.38 64.311 -
63.825 53.06 63As 63.925 8335 64.11 410 53.248 63.475 63.79 63.# 53.425 63.46 53.50 5343 63At 420 62.798 63.025 63.36 62.95 62.99 53.02 E065 53.09 53Js5 430 52.366 52.575 62.925 52.45 52.475 52.51 G2.64 62.56 52R5 440 61.921 52.125 62.42 82.065 52.10 62.14 52.175 . 52.21 . 52A1_ 450 ' 61.546 51.66 82.025 51A4 61A8 51.725 61.76 5126 460 &4.020 61.175 61.56 StA1
'* 51.175 51.22 61.25 61.30 61.50 470 $0.505 00.*s0 61.1 Si.14 80.7 80.74 80.78 00A25 51A35 400 50.00 90.20 50 42 80 86 80.22 30.265 50A1 80.35 505F5 40 49.505 49.685 30.13 00.175 40.714 4 .782 #A1 40A08 SOA00 900 48A43 6 .097 40A18 m.406 i
4 .152 40.203 40J64 S .306 48.86 S10 48.31 48.51 4 .06 40.101 I 48.46 48.515 48.67 48A25 48A8 #.?*5 4kO1 l 620 47A5 47A1 47A19 47.978 48.037 48.006 46.'155 48A5 630 47.17 47.29 47A6 47.362 47.428 47.494 47.06 47AD 640 46.51 47.23 47.296 46.658 46.726 46.794 46A62 46.93 47A7 850 45.87 46.80 46.068 46.142 46216 46.29 ASAS 660 45.25 45.92 45.994 45A2 45.30 45.38 45.46 45.54 45.82 4tA2 870 44.64 44.93 45.36 44.50 44.586 44.672 44.758 44A44 580 43.86 44.205 44.68 43.73 43.825 43.92 44.015 44.11 SSO 43.10 43.434 43A56 42.913 43.017 43.122 43.226 43.33 500 42.321 42.55 43.14 41.96 42.08 42.196 42.314 42.432 ' S10 41.49 41A16 42.283 40.950 41.083 41.217 41.35 41A83 620 40.552 41A4 ! S30 39.53 40.308 ' l 640 38.491 39.26 650 37.31 38A00 f l 660 36.01 3652 l 670 34.48 34A3R 683 32.744 32.'444 690 30.516 TABLE A.6 PROPERTIES OF WATER, DENSITY
- A.8
UNITW STATES h NUCLEAR REGULATORY C0tNH8860N nationa q {, - tei masusTTA sfasst u.w.,surfa sese ATLANTA,es0RetA 30833 p l
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 17 ?
~ ~
~~~~iU5E556 Uds5C5IU5dT TEEU5F5R d 6~ELUf6"FL U ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, H M ANSWER 1.01 (1.00) ;
C REFERENCE . - - N U S , -- V o l -- 4 , - p p - G --- - - --- - - - - -- - - -- - - - - - - - --- - J------- Nuclear Pwr Plant Operator Trng Prger HTFF and Thermo, Sect 2E CNTOr " Thermal / Hydraulic Princples and Applications, II", pp 10-43/44 ~ App At Pumps /Centrifuga1r Operating at Runout Conditions (3.1/3 1) ANSWER 1.02 (1 00) c
^
REFERENCE C:mprehensive Nuclear Training Operations (CNTO), pp 4-16/27 001/000; K5.13(3.7/4.0) ANSWER 1.03 (1.00) d REFERENCE CNTO, " Fundamentals of Nuclear Reactor Physics", pp 8-54/55 l 015/000; K5.06(3.4/3.7) l I ANSWER 1.04 (2.50)
- 1) Lower (Higher Stm Flow >> P sta decreases)
- 2) Higher (Less resistance to flow >> Other RCPs speed up)
- 3) Lower (Less total flow across core >> delta T increases, Tc goes down with rods in manual)
- 4) Higher (as above, delta T increases, Th increases)
- 5) Same (Primary power = secondary load)
REFERENCE NUS, Vol 4, Units 1.3, 3.2 - CNTO, " Thermal / Hydraulic Principles and Applications", pp 12-15/18 l 1 002/000; K5.01(3.1/3.4)
% UNITED STATSS NUCLEAR RE00LATORY COMMISSION h
... ,[ mg00006 N {. 101 MAAIETTA STAEET, N.W., SUITE 3000 ATLANTA, GEORGIA 30323 . D i I d
..... l 1
1 M " CTPLES OF NUCLEAR POWER PLANT OPERATIONr PAGE 18
--- isEER557sisiCi- sEsi isissiEE XE5 FEUi5 iE5s 7 ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, H M 8
ANSWER 1 05 (1.50) W
- 0) More Negative (+0.5 ea) '
b) Less Negative c) More Negative l REFERENCE Wastinghouse Nuclear Training Operations, PP. I-5.6 - 16 CNTO, ' Reactor Core Control *r pp 3-16/28 001/0003 K5.26(3.3/3.6) ANSWER 1.06 (1.50) o) Higher (+.5 ea) b) Lower-c) Higher REFERENCE TPT OP 1009.18 Plant Curves 001/0008 A2.07(3.6/4.2) ANSWER 1.07 (1.50) o) False (+.5) ) b) -Post accident heating of Reference Leg (+.5 ea)
-Reference Les leakage REFERENCE NRC IE Info Notice 84-70 (4 Sep 1984)
TPT Lesson Plan for Requal Cycle II-1985 011/0003 K4.03(2.6/2.9) ANSWER 1.08 (1.00)
- 1) Density difference between cold and hot les (+.5 ea) [F mwd j psk a/ M) l
- 2) Height difference between hot and cold less (or S/G and Core) l
wareseff.Tes i
. f.y g %'" NUCLEAR REQULATORY COMMISSION 9
,.! maamm -
2 101 MAMETTA STREET N.W 0WTE 3000 ATLANTA.GEOMMA 30828
%,*e..*
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 19 '
~ '
~~~~TUEEU66iUI55C5,~E55T TEI 5 FEE ~5U6~fEU56~fE6U -
ANSWERS -- TURKEY POINT 384 -86/02/03-DEAN, N N REFERENCE CgfD,,"#fdew ThermalM C. /Hydraul,$cPrinciplesandApplicationsII*r ba g c A 1.e58/h. pp 14-16/17 002/0003 K5.10(3 5/3 9) - -ANSWER -1.09- ( 2 . 0 0 ) --- - = c) Unit 4 (+.5) due to a lower Beta coefficient at EOL (+.5) b) Unit 3 (+.5) due to NTC being less negativer so Tavs must decrease seee to add + reactivity) (+.5) MC REFERENCE CNTO ' Reactor Core Control'r pp 3-21 & " Fundamentals of Nuclear Reactor Physics", pp 7-31 001/0008 K5.49(2.9/3.4) 1 K5.10(3.9/4.1) ANSWER 1.10 (1.25)
- 1) As the fuel burns oute less boron is required, which increases the boron worth (+.5)$ (or less boron means decreased flux hardening and a higher effective boron absorption cross-section so worth increases)
- 2) Fission products build up, decreasing _the boron worth (+.5)
(This is the overriding effect (+.25),[ REFERENCE SON /WBN License Cert Trns, ' Core Poisons *, pp 4 CNTO, " Reactor Core Control', pp 5-15/16,s7 - (g p fo rov , CWte' X for 7PT 001/000; K5.09(3.5/3.7) ANSWER 1.11 (1.50) - o) Amount by which core whould be suberitical (+.25) at hot shutdown conditions (540 des F) (+.25) if all control rods were tripped assuming highest worth rod fully withdrawn (+.25) and no changes in xenon or boron concentration. (+.25) b) No(already accounted for in SDH calculation) (+.5) REFERENCE l TPT T/S, pp3.2.2/3 i i TPT OP 1009.3 - l i I
f M ETLTM 7 NUCLEAR REGULATORY COMANSSION .l
.f.
, . t e1 asARIETTA 8TREET.N.W., SUITE 3000 ;y ATLANTA, GEO.MNA 30833
.l
\**ee* :
i
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATIONe PACE 20
~~~~ ~ ~
TU5EU66EUdUfC5,"U5dT TR 5F5R U6~iLU 6~FL6U ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, WM 000/0058 EK1.05(3.3/4.1) ANSWER 1.12 (1.00) This is due to " Uncovering
- of the sources by that bank (+.7) causing s' cpproximatealy_1/2 decade increase in count rate (+.3) -
REFERENCE
- TPT OP 0202.2, Step 4 2.2 001/0003 K1.05(4.5/4.4)
ANSWER 1.13 (1.25) At the end of 60 seconds: 32 SPM x 5 pan /in x 5/8 in/ste min = 100 pca (+.25) The reactivity addition rate: 5 pen /in x 32 S /8 in/ step = 100 pen / min ~ ' (+.25) The equation to be used is SUR = p+ p)/(Beff p) (+.5) The answer is .55 DPM ) REFERENCE CNTOr entals of Nuclear Reacstor Physics", pp 7-71/72 i EK1.02(3.6/3.9) l 1
\
l
90575 STATM
.~. ?
k NUCLEAR REGULATORY COMMISSION RE980N E 1
*l 101 MARIETTA STASET. II.W., SUITE 2000 g'l
-{ ATLANTA, GEORGIA 30823
% *e e* ,
s _ s PAGE 21 N 2.'__ ___________________________________________________ __ PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS t ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, W M CNSWER 2.01 (1.00) i c ; REFERENCE - ', TPT E0P E-1.2, 1.3, 1.4 004/0208- A4.02(3.9/3.8) - ANSWER 7.02 (1.00)
/b REFERENCE TPT SD117 'AFW', fig 7 TPr t ocre At W s+.<t 061/0001 K4.09(3.7/4.1)
WER 2 03 (1.00) REFERENCE TPT SD153 " Service a ire Water", pp 7-9 076/0003 K1.15(2.5/2.6) ANSWER 2.04 (1.00) o REFERENCE VEGP, Training Text, Volume 8, p. 16b-3 and Fig. 16b-1 &6 VCS, GS-2r Safesvards Power System, pp. 27 a 28 TPT DWG 5610-T-E-1592 062/000i K4.09(2.4/2.9)
meo y, SITED STATBS
, + NUCLEAR RE ULATORY COMMISSION . . .E .
f( a{ 101 MAA8ETT A STREET.N.W,Sufft 2000 ATLANTA GEOA04A 30328 - d
*e e* -
?
i
?, *8 ANT DESIGN INCLUDING SAFETY AND ENERGENCY SYSTEMS PAGE 32 .
_______________________________________________________ v ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, W N ; ANSWER 2.05 (1.00) c) True (+.5 as) b) False REFERENCE . TPT SD155
- Plant Air Systems *r pp 6/7 078/0003 K4.02(3.2/3.5)
ANSWER 2.06 (2.50) I o) 3 b) 5 c ) .W i d) 6 s a) 2 (us er *V)' o r %(vun t) REFERENCE Fctley SDr 'RCS'r Fig 7 .. NA NCRODPr 'RCS*8 *ESF-ECCS*l "CVCS't "RHR' TPT SD7 'RCS*r pp 65-67,' Du>4 (5'0* T *f- V fD# 002/000; K1.06(3.7/4.0)I K1.09(4.1/4.1); K1.08(4.5/4.6) i ANSWER 2.07 (1.00) Backup to AFWi Unit 4 Condensate; Demin Water Storagei FWRV Bypass (+.25ea) REFERENCE TPT Requal Cycle IV-1985, Day 3 TPT SD112 " Condensate and Main Feedwater', pp 6/7 059/000; K1.03(3.1/3.3) ANSWER 2.08 (1.25) a) 5 (+.25) b) RHR systemi RCPs (+.5 ea) REFERENCE TPT SD13 'CVCS', pp 46-51 1 1 1 1
% UNITED STATES NUCLEARRE ULATORYCOMMISSN)N
{.
\'" . . J$ =0.ON .
141 MARIETTA 8TMET,N.W., Sufft 3905 - ATLANTA, GEOA04A 30833
'. * *t nESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 23
}
ANSWERS -- TURKEY PDINT 344 -86/02/03-DEAN, W H 004/0208 PWG 7(3.4/4.1) . ANSWER 2.09 (1.00) o) Prevent reverse flow from the VCT (+.5) , b) Prevent lifting ti seal and having it hang up (+.5) REFERENCE TPT SD8 'RCPs", pp 27-28 003/0008 PNG 7(3.5/3.9) ANSWER 2.10 (1.50)
- 1) Iodine isotopes are more readily maintained in solution (+.75 ea)
- 2) General corrosion rates of structural material is reduced REFERENCE TPT SD25 " Containment Spray *, pp 11 026/0003 K4.02(3.1/3.6) t,N':"C P 2.11- (2.00) -h,j,f,,QO
- 1) Max Fuel Element Cladding Temp < 2200 Des F (any 4 of 5 at +.5 ea)
- 2) Cladding Oxidation < 17% thickness .
- 3) H2 generated by Zirc-H2O reaction < 1% of max possible
- 4) Core remains in a coolable geometry !
- 5) Provides for long term decay heat removal j REFERENCE '
10CFR50.46 TPT SD21 'ECCS's pp 5 006/050; PNG 4(4.2/4.3)
g me. DesaT50 STATS 3 g a
- 8. \ NUCLEAR RECULATORY COMMISSION RE040N N
'$* 101 MARIETT A 8TREET.N.W SulTE 2000 ATLANTA, OE044MA 30333 -
*'\,..... f z
9
?. *l. ANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 24 -
ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, N M . l l 1 2 ANSWER 2.12 (1.00) prevent water from flashing on the shell side of the regenerative Nt Exengr t (caintain RCS pressure in the Nt Exchgr with the high temp water) (+.75) - cnd prevent damaging RV-203 (LTON Relief) (+.25) REFERENCE SONP System Descripe 'CVCS'r pp 9 . TPT SD13 'CVCS*r pp 17 004/0108 K4.03(3.1/3.6) CNSWER 2.13 (1.25) c) Backleakage from S/Gs via check valves (+.75)
- b) Vent the pump casing ~(+.4) once a shift (+.1) -
REFERENCE TPT EO 63, Cycle I Requal-1985 035/0108 K1.01(4.2/4.5) (I;1.u)Gi ANSWER 2.14
=1 b) d==o=* + = rd i c r. . . . . u i u ; w . s1 .ai A#@
Accordian type mechanism w/ a fusible leak (+.5) if hot gases pass thru ducte fusible link melts (at approx 165 des F) releasing the damper and sealing the duct. (+.5) REFERENCE TPT Lesson Plan for Requal Cycle IV-1985, ' Appendix R Update' 086/000i A1.04(2.7/3.3)
i g "88g INETED ST' TBS j
, .# ',g" NUCLSAR REGULATORY COMMISSION naosoNu y
. . . , {8 101 IAAAIETTA STREET N.W.,8047E 2000 0l ATLANTA, oEORGIA 30333 ~~
s,
-1
- 3. INSTRUMENTS AND CONTROLS PAGE 25 4 u
ANSWERS -- TURKEY POINT 384 -86/02/03-DEAN, W H ANSWER 3 01 (1.00) . b
~
REFERENCE '_t - TPT SD117 'AFN', pp 10 013/0003 K4.04(4.3/4 5) ANSWER 3.02 (1.00) b REFERENCE TPT Less an on ' Gamma Metrics Neutron Detector' Requal Cycle IV-1985 008 K6.01(2 9/3 2) ANSWER 3.03 (1.50)
- 0) [ (+.5 ea) b) 2 c) 1 REFERENCE P
Sg1 ggs", Fig 11, 12, 13 035/010; K4.01(3.6/3.8) i i
\
ANSWER 3.04 (1.00) o) Arm only (+.5 ea) b) Arm and Actuate REFERENCE TPT SD105 ' Steam Dump System', pp 9-16 ; 041/020; K4 14(2.5/2.8), K4.18(3.4/3.6) ) l
% #g UNITED STATES NUCLEAR REOULATORY COMMISSION q
.'.,,8 nEeION N j
,g 101 MARIETTA STREET, N.W., SulTE 2000 j ATLANTA,0 EOR 04A 30383 .
\*e*** .
t . .8 TMRTRUMENTS AND CONTROLS PAGE 26 ( ANSNERS -- TURKEY POINT 314 -86/02/03-DEAN, N M ANSNER- 3.05 (1.50) b o) False (+.5 es) t b) False- - I c) True REFERENCE
.TPT"SD63 'ESFAS', pp 32iS2, FIG 14 ,
006/000; K1.02(4.3/4.6) ' ANSWER 3.06 (2.00) o) 5 (+.5 es) b) 4 e) 1 :. d) 5 (vast 3) or 6(wwr REFERENCE - TPT SD5 ' Rod ntrol System *e Fig 12 yPr mar, r4to- , Ides / /7 ' , 001/0503 K4.01(3.4/3.8) ANSWER 3.07 (1.00) Lovel III Channel (+.5) failed high (+.5) REFERENCE Westinghouse PWR Systems Manual " Primary System Control", pp 12-14 TPT SD9 'PZR and Pressure Relief *, pp 38-40, 57; DWG 5610-T-D-15 011/000; A2.10(3.4/3.6) ANSWER 3.08 (1.50) gg } e) Loss of voltage sensed by 2 relays on either 4KV bus A or B (+.5)[#[frN;O gy. (ef b) -Hanually reinitiate the SI signal upon the loss of voltage (+.5 ea) y
-Manually restart SI equipment taking care not to overload diesel REFERENCE TPT SD170 ' Emergency Bus Stripping / Load Sequencing', pp 9, 16-17, (,
OwCe rg to. 7. c - t, f t G
usefes STATES ?
, T p %'g ' NUCLEAR RESULATORY COMMISSION Mmmau i{8 ist asAmeTTA stasettu.w., susTE asco ATLANTA,GEOROLA 34423
\,*ene* .
C 1
~ ~ ' ~ ~ ~ ~ " ~ ~ ~
- 3. INSTRUNENTS AND CONTROLS PAGE 27 f
~
ANSWERS -- TURKEY POINT 384 -86/02/03-DEAN, W N 064/0008 K4.10/K4.11(3.5/4.0) -
,, . E' i ANSWER - 3.09 .(1.20) .. --
p IRM High Flux Reactor Trip (+.4 es) d 1.
- 2. IRM High Flux Rod Stop -
j
- 3. PRM High Flux Reactor Trip - Low Setpoint '
- '~
REFERENCE Cote SD-IPE, p.'19 ' SONP PLS pp 5 - l TPT SD4 *Excore NIS*r pp 76 015/0008-K4.07(3.7/3.8) ,
~
ANSWER 3.10 (1.80) . ' -PZR high water level ( + . 3 e a-) i-
]
/ -PZR lo pressure
/-Lo prieary coolant flow -
.-'~
-RCP breakers open (two pumps)
/ -Under voltage on both 4KV buses . / -Turbine trip REFERENCE SONP PLS, pp 4 TPT SD63 'RPS", pp 37 012/000; K4.06 (3.2/3.5) ANSWER 3.11 (2.00)
- 1) OMS Mode Control Switch in ' Low Press" (+.5 ea)
- 2) PZR PORV Control Switch in ' Auto'
- 3) PORV isolatioin valve (535 or 536) open
- 4) Power available to PORVs and PORV isolation valves REFERENCE TPT SD7 'RCS*, pp 49 002/000; K4.10(4.2/4.4)
h fg .MM n:T.
. */ g' NUCLEAR RESULATORY COMMISSION s * ,) MenON E
{ 101 MAhlETTA STMET.N.W., SWITE 3900 ATLANTA,OE0AelA 30823
\..... C
.. t
_I__________________________
-86/02/03-DEAN, W H I ANSWERS -- TURKEY POINT 314 1
C.NSWER 3.12 (1.50) , i
'1) Hi side of S/U Xfrar energized (+.3 es) e Unit 3 S/U Xfrar has been bus d1 eared
~
ft
- 2) *
- 3) DG Breaker to 4A bus open *
- 4) 4A to 4B bus tie open 5)- 4A-bus-not-locked-out Wg gtk REFERENCE y0P - f TPT SD170 ' Emergency Bus Stripping / Load Sequencing's pp 6-7 [ h T M 000/0558 EA2.03(3.9/4.7)
ANSWER 3.13 (1.00) cQTd
- 1) aration) you could parallel HCCs on different units' (eliminate unit sep(M)--(' ."n :
- 2) the MCCs could not be in sync if paralleled causing circulating. currents ~ '
and create overload conditions g) REFERENCE TPT Lesson Plan 20-OL, APP B " Replacement of 120 VAC Inverters *r pp 5 hY 062/000; Y'.T N~-lY A3 04(2.7/2.9) l I
a#
/ WINTSD STATSS NUCLEAR REGULATORY COMMISSION
. . . ,8 pee 00N 5 101 blARIETTA STREET.88.W., GUITE 2000
'{ g ATLANTA,080#elA 30828 g
.. A
- 4. . PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 29
--- EK515E55i53C 55 EYE 5E------------~~-----~~---
1 l ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, W M ' ANSWER 4.01 (1.00) o - 3 REFERENCE ^ - - l TP T-EOP-FR-S .1 001/0108 A2.08(4.4/4.6') ,,_'. ANSWER 4.02 (1.00) g REFERENCE , 10CFR20.5 PNG-15! Radeon Knowledge (3 4/3.9) vr ANSWER 4 03 (1.50) , c) False (+.5 ea) b) True c) False REFERENCE Wastinghouse User's Guide for TPT E0Ps, pp 5-12 ANSWER 4.04 (2.00)
- o. Equalto//fW /kau (0.5)
- b. Fast (0.5)
- c. K CHw l1G (0,5)
- d. Governor control (0.5)
REFERENCE Cat, OP/0/B/6350/11, enclosure 4.4, pp. 1 -2 TPT OP 4304.1, pp 6/7 [fk'sk'h.0, pp 7/8 064/000; A2.03 and A2.09 (3 1/3.3)
g UNITS STATES j 4 NUCLEAR REGULATORY GOtHA18SNl)N < [ . . ,,,[( naesoNu 101 blARIETTA STRSET, N.W., SUITE 2900 5 ATLANTA, GEORGIA 30823 3 3 9
~
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 30
~~~$R5656L 656AL 66Eri6L -
l
-86/02/03-DEAN, W M ANSWERS -- TURKEY POINT 314 ANSWER 4.05 (1.00) a c) 10EE(-8)3 1 (+.25 es) - -
?
b) 1 - - - . g c) 1000 , REFERENCE . j TPT OP-0202.2r pp 2/3r 18 j
' 4 001/0508 PNG-7(3 6/4 1)
ANSWER 4.06 (1.50)
- 1) ,NTM Air Particulate Monitor ( Any 5'Yor +.3' ea) -
- 2) CNTM Radioactive Gas Monitnr '
i
- 3) Component Cooling Liquid Monitor
- 4) Condenser Air Ejector Gas Monitor
- 5) S/G Liquid Sampl#e Monitor
- 6) Plant Vent Radiation Monitor .
REFERENCE TPT ONOP 1008.2, pp 2 000/0281 EA1.06(3.3/3.6) ANSWER 4.07 (1.00) 7M Containment pressure (+.35) > 4 psis (+.15) Containment" radiation (+.35) >'DOEE5 R/Hr (+.15) REFERENCE 49MP O\1) > tN *r o
- o g)
Westinghouse background info for TPT E0Psg " Instrumentation Accuracy'epp 11 022/000; K3.02(3.0/3.3) %t4tw L%% for emP G wra% # pp tg y ( I i {
r I~ .
- l. p% UNefensTATES
, .d(
.', s 5 h ~
NUCLEAR REGULATORY COMMISSION n9980N ll
\
a { 19118AAIETTA STREET, N.W 8UITE 3900 ATLANTA, GEORGIA 30333
*4 d
*eee*
t
- 4. PROCEDURE 8 - NORMAL, ABNORhAL, EMERGENCY AND PAGE 31 b
--- iK5i5C55iEXE 55EfE5C------------------------
1 ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, W H 1 ANSWER 4.08 (1.50) [g,yg g fjg) ;
- 1) ' Manually trip reactor (or open trip breakers locally) D ." ;;)
- 2) mggy 4 ; y v/ g eg g gg
$~) pEvacuate Manually CRtrip andturbine
,:r"_
E; .....,1 3 ..:' 9; tic: ;.. rt- ahment *-- , ;.it 9"a repor+ Tur&W Cpers4er ikule ~[f. 3) ~ ~~ ~ -^ ~ - - REFERENCE .- , TPT 0-ONOP-103, pp 5 [/.#r/boi ////6&T 000/0688 PNG-11(4.5/4.5) l ANSWER 4 09 (1.50) PZR PORVs Closed
-~
- 1) (+.5 ea) . - . .
- 2) LTDN Isolation Valves Closed
- 3) ' Excess LTDN Isolation Valves Closed REFERENCE TPT E0P-ECA-0.0, pp 3 000/0568 PNG-11(4.5/4.6)
ANSWER 4.10 (1.50) Core Exit T/C (+.5 ea) T-Hot RCS Subcooling l 1 REFERENCE SON ES-0.3, pp 6 l TPT ES-0.2, pp 5 EPE-074; EA1.02 (3.9/4.2) l
ya yleTW STATES A( ' NUCLEAR REGULATORY C00HNSS10N r a e:, { Reeloesa it1 MARIETTA STREET, N.W.,9UITE 2900 g g
- ATLANTA,880A04A 80888
\*e.** $
5 f JN FDURES - NORMAL, ABNCIRMAL, EMERGENCY AND
- f. . PAGE 32
~ ~ ~~E5656L6556AL C6NTRUL~~~~~~~~7~~~~~~~~~~~~~~
ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, W'M & i
, ,\
ANSWER , 4.11 (3.00) _ . . . c) -SI' Annunciator DN ~~ (+.3 es responr>e) -
-SI pumps running ,
-8tN.Rausps running. J
-EDGs running - ?
b) -CNTMT Purge / Supply fans OFF J.; -
~
a
-Purge Valves CLOSED y
-Instrument Air Bleed Valves CLOSED
-Verify Control Room Ventilation Isolation ..
c) -AFW steam supply MDVs OPEN
-AFW Flow Regulator Valves OPEN REFERENCE ~
TPT E-0, pp 4/5 " 000/0073 PNG-11(4.4/4.5) ANSWER 4.12 (1 50) - To prevent excessive depletion of RCS inventory (+.5) such that the RCP trip occurs (+.5) at a point where the break would completely uncover the core (+.5) REFERENCE Westinghouse background info for TPT E0Psr 'RCP Trip / Restart' 000/0098 EK3.23(4.2/4.3) i
.., . a s v o *
- ENCLOSURE 4: Utility Comments P ge 1 NRC Exam Secti:;n 1
,c** QUESTION RO 1.03: (SRO 5.03, SRO Requal 5.02)
'l When performing a reactor S/U to full power that commenced five hours after a trip from full power equilibrium conditions, a 0.5%/ min ramp was used. How would the resulting xenon transient vary if instead a 2%/ min ramp was used?
- a. The xenon dip for the 2%/ min ramp would occur sooner and the magnitude of the dip would be smaller.
- b. The xenon dip for the 2%/ min. ramp would occur later and the magnitude of the dip would be smaller.
- c. The xenon dip for the 2%/ min. ramp would occur sooner and the magnitude of the dip would be larger.
- d. The xenon dip for the 2%/ min. ramp would occur later and the magnitude cf tha dip would be larger.
e
RESPONSE
Y We request that answer "a" be accepted as the correct answer because the xenon dip would occur sooner and the magnitude of the dip would be smaller for the 2%/ min. ramp rate. , REFERENCE 2 Westinghouse Simulator Training Book - Reactor Theory and Core Physics Page I-5.76. l l l
.R 1-5/dj l I
02/06/86 l P;ge 4 NRC Exam Section 1 j' ^,
.- QUEffflON RO 1.18: (RO Requal 1.10)
There are two effects that cause differential boron worth to change over core life. List these two effects, their relative impact on differential boron worth and indicate which effect is the overriding factor.
RESPONSE
We request that the decrease in boron concentration as the overriding effect over core life also be considered acceptable. The boron worth value used at PTN is determined by use of an integral boron worth curve and a boron worth correction factor curve. The integral boron worth curve is a grapn of the total reactivity worth for any boron concentration assuming a poison free reactor. The correction factor curve is used to get a correction factor for the integral boron worth based on a total amount of poisons in the reactor. Each of these graphs, as
,'- well as the Xenon and poison worth graphs, have 3 curves on them for BOL, MOL, and
- EOL.
To determine the differential boron worth from this data, the following sequence would be followed.
- 1) Determine the integral worth for a given boron concentration
- 2) Determine the correction factor for the total Xenon and Samarium in the reactor for the desired conditions.
. 3) Divide the integral worth by the correction factor to determine the corrected poison worth.
- 4) To get the differential worth, divide the corrected poison worth by the given boron concentration.
The above method will give the differential boron worth for any point in core life. The method described above can be used to determine the overriding cause for the change in differential boron worth. The altered method would be as follows:
.R1-5/dj
02/06/86 l
- ?
P:ga 6 l NRC Exam Section 1 ' QUESTION RO 1.17: (RO Requal 1.11) (a) What is the definition of Shutdown Margin (SDM)? (b) If a stuck rod exists while the reactor is at power, what adjustment, if any, must be made to the SDM calculation?
, :J4 RESPONSE:
, M,,,. .: , We request that " shut down margin should be increased by boron addition to compensate for the withdrawal worth of the inoperable rod" also be considered an acceptable _ _ answer.
REFERENCE:
Turkey Point Unit 3 and 4 Technical Specifications section 3.2.4C.
,e
, - -w h
. R 1-5/dj
l 02/06/88 Pega 8 NRC Exam Section 1 QUESTION RO 1.22: (RO Requal 1.13) i 1 Unit 4 is just critical in the intermediate range when rod D-4 (which was at 140 inches) begins to withdraw at 32 steps per minute. Assuming a differential rod worth of 5 pcm/ inch, what is the SUR 60 seconds into this rod withdrawal accident? Show your calculations.
, RESPONSE:
4. We request that this question be deleted. The basis for this is that the formula necessary to solve this question was incorrectly stated on the formula sheet provided with the exam. The formula sheet equation was: SUR = 26 *, + l [-p see attached
.s
. ]
. R I-5/dj
Lp/Uu/96
.- : . Page 2 l l
NRC Exam Se:;tirn 2 l
. QUESTION RO 3.03: (RO Requal 3.03)
Which of the statements below regarding Unit 3 APW pump steam supply valves on a loss cf the 3A 4KV bus voltage is correct?
- a. Steam supply valves from all 3 8/Gs open.
- b. Steam supply valves from A and B S/Gs open.
- e. Steam supply valves from B and C 8/Os open.
- d. Steam supply valves from A and C 8/Os open.
- e. No steam supply valves will open a..-m,~ s it t kes loss of voltage on NthIEV two 3A and 3B to cause the valves to open.
RESPONSE: ' We request you accept answer "b" as.the correct answer,.as less of volt _ age ,on_4EV bus _ _ _ __
"A" opens steam supply MOVs from "A" and "B" steam generators. This feature is common to units 3 and 4.
C;. ) marsanNcs: TP-5610-T-LI, Sheet 15 i l
- RI-5/dsp l
#2/04/84 Page 3 NRC Exam Section 2 QUESTION RO 2.98: (RO Requal 2.04) ,
Match the RCS penetrations in~ Column A with the appropriate RCS loop segment listed in Column B. (column B ltems may be used more than once but only one response per penetration) Column A Column B
- a. Excess Letdown 1. 1) Loop A cold leg
- b. Par surge Line 3) Loop A hot leg
- e. Alternate Charging 3) Loop A Intermediate leg
- d. Par Spray Line 4) Loop B Intermediate leg RHR Suction - --~ - - Loop B hot leg
~ ~ ~ ~ - ~ ~ ~ ~ ~ - - - ~ ~ - - ~
- e. -- ~ ' ~' 5) -
- 8) Loop C cold leg f) Loop C hot leg .
n---. - - - . - . - . - . . . . . . . . .. . .. .. ..
RESPONSE
Part C
@ We request you accept answer #7 for part e of this question as alternate charging is ! associated with Loop C Hot Leg.
Part E We request you accept answers #2 or #7 for part e'of this question as units 3 and 4 differ in this respect. Answer #7 or C Hot Leg applies to unit 3 and answer #2 or Loop A Hot Leg applies ot unit 4. REFERENCR
.Part C TP-5610-T-#-4501 Part E i TP-5610-T-8-4510
, *R1-5/dsp
02/06/86 P g2 6 NRC Exam Section 2 QUESTION RO 2.20: (RO Requal 2.14, SRO 8.19, SRO Requal 8.12) a) What indication does a control room operator have that a fire damper has actuated? b) Explain in detail the design features which allow a fire damper to auto close when hf.
%N g required.
RESPONSE
; __ ____._._.1_._ . . . _ _ . __ _ . . . _ . _ _ - . - _ _ _ _ _ _ _ _ . _ _ _ _ _ _ - - _ . _ _ _
We request this question be deleted for the following reason: The fire dampers are a portion of the larger appendix "R" modifications being installed
~
at Turkey Point. The installation is not complete and has not been turned over for plant use. A training overview wa presented to licensed operators as "Look Ahead" to inform g.4 them of the appendix "R" modifications in progress. The presentation given to the
~
i Group X students did not cover the fire dampers) Mme instruction given was not detailed because the modifications were still in progress. More detailed training will be provided when the appendix "R" modifications are completed and ready for turnover.
REFERENCE:
t
- Lesson 82-OL Appendix A page 4 titled Appendix "R" Mods i (Presented to Group X Students)
I L 1 "R1-5/dsp
L- %2/06/44 i Page7 h@: \"* NRC Exam 8ectt:n 2
- Y QUESTION RO 2.21: (SRO 6.20, SRO Requal 8.13) l
?b *
/ a) Describe the runback process that occurs with the Main Turbine when the OT Delta T setpoint is exceeded? b) If the Power Range " ROD DROP AUTO TURBINE RUNBACE" is bypassed, what conditions must exist and what system will initiate a turbine runback? RESPONSES
;j. ' ' '
Part As __e._____..._ -__ _ __ _.. . . _ - . _ _ _ _ _ _ _ . _ _ _ . . _
' ~
We request the NRC answer [ey corrected to read as follows:
. _ _ _ . "a) Turbine ls runback at 200%/ min (Unit.3158% miti.) for 1.5 sec.,l stops for 30 sec. _
then repeats cycle if condition still exists." Part B: We request that the portion of the answer "... what conditions must exist..." be deleted. The condition "... Reactor Power is 70% as sensed by turbine 1st stage impulse pressure." is incorrect for the situation addressed in the question. REFERRNCE: Part A Dwg 5610-T-L-1, sh. 21 Part B Dwg 5610-T-L-1, sh. 21
*It1-5/dsp
02/06/86 Paga1 NRC Exam Section 3 h ys QUESTION RO 3.03: (RO Requal 3.02, SRO 8.05,8RO Requal 6.04) Which statement below correctly describes operation of the GAMMA-METRICS Neutron Flux MonMor in gamma flux fields between 10,000 and 1,000,000 R/hr (ie. high radiation j fields).
- a. The monitor is not designed to operate in such high level radiation fields.
f "3 b. The monitor will operate satisfactorily in these radiation levels, but an. . l adjustment should be made to discriminate against the higher gamma flux. l
- c. The monitor will operate satisfactorily, but the output signal from the detector will not increase linearly due to lack of voltage saturation in the detector.
- d. The monitor is designed to operate as well in this high a level gamma flux as
. It does in much lower radiation fields. )*,y
RESPONSE
We request this question be deleted for the following reasom The Gamma-Metrics Monitor is a component of the larger safe shutdown system. The Safe Shutdown System has not yet been turned over for plant use. A partial turnover of the Gamma-Metrics Monitor was performed in 1985. The Gamma-Metrics Monitor was e functional at that time but was to be used for indication only. Procedure changes and training were determined to not be necessary. Full training will be implemented at time of Safe Shutdown System turnover. A training overview was presented to licensed operators as "look ahead" to inform them of the new instrument in their control room. l However, the instruction given was not detailed because of the limited purpose of the Gamma-Metrics Monitor at that time. l l
*RI-5/dsp
~
Q2/06/86
.- e Page3 NRC Ex*:m Se!,ti:n 3
,g QUElfflON RO 3.04: (RO Requal 3.03)
- u. ..
Indicate whether there are 1, 2 or 3 SELECTABLE detector inputs for each of the following parameters attilzed by the S/G Water Level Control System. a) 5/G Level b) Feed Flow e) Steam Pressure
RESPONSE
We .are . requesthig that you . accept _2_ (two) as the. number of selectable .8/G _ level ___ detector inputs to the' 8/G 1evel control system in part "A" of the question. The SD11 (Fig.11) reference shown on the answer key has not been updated since the second _ _ _ . _ _ _ . . . _ ehannel was added. _ _ _ _. _. _ _ _. _ _ _ _ _ _ _ _ _ _ ' _ _ _ _ _ _ . RRFERENCE: Plant drawing 5610-TD-17 l l l 1 i 4 l
- R1-5/dsp
~
Q2/06/86
.- Page 3 NRC Exem Secti:n 3 QUElrl'lON RO 3.08: (RO Requal 3.06)
(]cp . Match the interlock descriptions in Column A with the appropriate logic required to cause rod withdrawal to be blacked in Column B. (column B ltems may be used more than once) Column A Column B a) Power Range High Flux 4103% power 1. 1/2 b) Overtemperature Delta T rod stop 2. 3/2 e) Intermediate Range High Flux 3. 1/3 d) Power Range Rod Drop 4. 2/3
-_ _ _ 5. _ 1/4 - _ _ _ _ . _ _ _ _ _ _ _
6.' 2/4
- 7. 3/4
RESPONSE
We are requesting that "5 and/or 6" be accepted as correct answers to part "D" to this
'~
question. The Power Range Rod Drop logic is currently different for Units 3 and 4. The correct answer for Unit 3 is "5" (1/4). The correct answer for Unit 4 is "6"(2/4). As the Unit number was not speelfled in the question, either or both of the above should be acceptable.
REFERENCE:
Plant drawing 56. ? '..-1, sheet 17 j See Note 1 l
- RI-5/dsp
(/ ,', Q8/eW/c6
.y Pcge 4 NRC Exam Secti:n 3
,.. QUErrION RO 3.12: (RO Requal 3.08, SRO 6.10)
W.' . a) Give the location and the number of UV relays that must be energized to initiate bus stripping on a Loss of Off-Site Power. b) To ensure needed vital equipment starts on a Loss of Off-Site Power following an SI that has been RESET, the operater can perform what two actions?
RESPONSE
3 We request you accept the following additional answers for Part a of this question. 2 Loss of voltage relays per bus (physically located inside the sequencer cabinets) Under voltage relays on the associated 480V load centers. REFERINCE: SD-170 page 6 Dwg. 5610-T-L-1 sh.13 i
- R 1-5/dsp
02/06/86 Page5 NRC Exam Sesti:n 3 QUESTION RO S.17: (SRO 6.14, SRO Requal 6.09) o List the 4 sets of ECCS related valves required to mitigate a LOCA which have their j l control power breakers racked out during critical operations. l
RESPONSE
We are requesting you accept valves 863 A & B as an additional correct answer. These valves are required to be controlled by T.S. 3.4.1.a.7 on page 3.4.2. , 1 i l i i l l t i i
- R 1-5/dsp
Pcga8 NRC Exam Secti:n 3 1
; ., QUESTION RO 3.21 (RO Requal 3.13, SRO 8.23, SRO Requal 8.14)
Qb - The Alternate Source Transfer Switches associated with the recently installed 110 VAC inverters have key locks to prevent 2 switches cf the same channel being selected to ALTERNATE at the same time. What are the purposes behlixt this administrative key controlf
RESPONSE
Please accept as an additional correct answer for part 1 of this question:
"You could parallel the CVTs on different units." _ _ _ _ , _ _ _
REFERENCE:
5810-T-E-1592 - l l *R1-5/dsp I
l 02/06/88 , Page 1 i NRC Exam Sectlen 4 ; 4:i. . QUESTION RO 4.84: (RO Requal 4.84) f
- For the following paragraph, choose the correct words from the options given after the paragraph that correctly complete each blank.
When paralleling a diesel generator to the grid, the generator voltage should be A - the line voltage. The diesel generator is synchronized to the grid by observing the synehro pointer as it moves slowly in the h direction and closing the generator breaker when the potat is a the vertleal position. The power (MW) output of the generator is then raised by adjusting the d. Chose from the following: - l a. Iower than / equal to / higher than
' _ . - _ _ . - _ . .- __ b. ..- - slow / fast - - - - - - - - - - - - - - - - - - _ - - - - - - - - - - - - - - - - - - - - - - - .
- e. 5 minutes to / at / 5 minutes after
- d. governor control / voltage regulator / stator cooling b
t -- We request that either " equal to" gr " higher than" be accepted as correct answers for part "A" of this question. Adjusting incoming voltage slightly higher than running voltage has been stressed in operator training to ensure a lagging power factor after breaker closure.
REFERENCE:
SD 137, page 44 1
- R 1-5/dsp
02/06/86 J PCgo 3 l 1 NRC Er.am Section 4 QUESTION RO 4.09: (SRO 7.08, SRO Requal 7.05) (-[} List 5 possible alarms (setpoints not required) on the Main Control Board that would be ; indications that an inadvertent dilution were occurring while the Unit was at power. i (Assume Rod Control is in MANUAL, NO Rx Trip occurs an NO operator actions are i taken to mitigate the dilution). , l RESPON8R _ Due the lack of speelfle information with regard to initial conditions, this question lends itself to many possible answers. 1 e. We submit the following list of additional alarms which may be indicative of an j Inadvertent dilution and request they be included in the answer keys
._ -.~ ...- ..... . - . . . . . -. .- .
F, q' I
- 1 i
- R I-5/dsp
1 7 92/06/84 j 9] Page8 i c ' NRC Exam Sectirn 4 l QUESTION RO 4.12: (RO Requal 4.07, SRO 7.09,8RO Requal 7.06) f/(M ! l List the two conditions (including setpoints) which determine Adverse Containment ) conditions.
RESPONSE
We request that 1800F be considered an acceptable answer. The background information for PTN's EOP's wars based upon the standardised , Westinghouse Owners Group bases. However, as plant speelfle data was considered it
-~-'was determined ~ that the most ~ limiting ~1nstrument would ~ require ~ that ' adverse ~
containment conditions be either 1800 F or 1.3 X 105 R/hr. l
-RRFERENCE: - - - ----- - - - - - - - - - - - - - - - - - - - - - - --
-- - -- - - -- -- -l Westinghouse setpoint study for EOP Generation, pg.167 i
i
- R1 -5/dsp
02/06/86 Pcge 7 NRC Exam Section 4 h QUESTION RO 4.13: (RO Requal 4.08) What are the Unit 3 RCO's immediate actions if the word is passt d " Fire in the control room, shift personnel report to assigned control room evacuation st stions"?
RESPONSE
a We request that the answer for this question be changed to read: tM sg. .
- 1. Trip the reactor
- 2. Trip the turbine _ _ _ _
- 3. Assist the PS-N
- 4. Evacuate control room - report to turbine operator shack
REFERENCE:
*-ONOP-103, page 5, section 4.4 <i,.in.
v,ys;- 1 ~~
,f5 [*
1 1
- R 1 - 5/dc l
l
02/06/86 Pcga 9 NRC Exam Section 4 k.~. f*;g;l'? e 7 QUES 110N RO 4.19: (RO Requal 4.12, SRO 7.21, SRO Requal 7.12) During a small break LOCA (SBLOCA), it is required to trip the RCP if the trip criteria are met. If forced flow through the core promotes cooling, why are the RCPs tripped?
RESPONSE
We request that you modify your answer to state the following:
"The reason for purposely tripping the RCPs during accident conditions is to prevent excessive depletion of RCS water inventory through a small break in the RCS, which might lead to severe core uncovery if the RCPs were tripped for some reason later in the accident."
^ ~~ ~ ~~ ~~ ~ ~ ~ ~
~The basis for our request is t' hat this will reflect the actual wording of the reference ~
stated in the answer key. Qiv
REFERENCE:
Turkey Point Emergency Operating Procedures Student Information Book Section 7, page 8
- R I-5/dsp
02/06/86 P ge 10 NRC Exam Section 4 k QUEFFION RO 4.20: (SRO 7.22, SRO Requal 7.13) a) What indication is used in the procedures'to denote sub-tasks which must be performed in sequence?
RESPONSE
We request that you accept as an additional correct answer to this question to be (g((b, l'y "nu mbers". The basis for this is found in the document referenced in the NRC answer key, Westinghouse User's Guide, page 3-3 which says; If sequence of performance is important, then the sub-tasks are designated by letters (or
~~ ~ ~ ~ ~~~ ~ ~
' ' ' ~ ~ ~ ~~~numbirs if finer detail is provided)'~~ ~ ~~~ ~ ~~~ '~ '
- f. -
REFERENCE:
Q.s.5 Westinghouse User's Guide section 3, page 3 I 1 1
- R I-5/dsp
92/06/86 Page1 NRC Exam Secti::n 5 QUESTION SRO 5.13: (SRO Requal 5.10)
)
What are the four conditions that Tech Specs say must be met to ensure the Nuclear Enthalpy Rise Hot Channel Factor is maintained within limits during periods between in-core survelliances?
RESPONSE
We are requesting that you accept 12 steps as another acceptable answer. The basis for this is that the Technical Specifications bases state that maintaining the rods within steps precludes - a rod - misalignment of greater than 15 inches - with - - consideration of maximum instrument error. _ _ _ . REFERENCE __ _ _ _.__.- _ _ . - . ,____ _ _ _ . _ _ __ _ _ T.S. B3.2, Pg. 3.2-5 O l
- RI-5/dsp
- - --,,.--,---gv----+---ey e- - , -w-g -
3 ---u e--. - - - - --
02/06/86 P:ge1 NRC Exam Section 6
- k. QUESTION SRO 4.11: (8RO Requal 6.06) ,,
Fill in the blanks in the statement below regarding the spent fuel pool holst: The hoist and bridge controls are interlocked to prevent raising or lowering the load while the is moving. If the upper limit position switch falls to stop a load lift, the will stop the hoist a few inches higher. The upper limit position switch is set so that the bottom nossle of the fuel assembly will clear . .
RESPONSE
We are requesting you delete part 2 of this question. This portion of the question is in error because SD-44 is la error. The upper position (lever) limit switch is designed for emergency eut-off only. It is not intentionally used as the normal stopping device. The geared upper limit switch will be 1 () used in normal operation to limit the maximum height to which the' hoist can be operated. This is documented in OP-16304.1; spent fuel pit bridge crane-periodic test. On part 3 of this question we are requesting that you accept the upender as an
- additional answer because it is the highest obstruction in the spent fuel pit and canal.
*RI-5/dsp
Q2/06/86 P ge 1 NRC Exam Secticn 7 QUESTION SRO 7.14: (SRO Requal 7.09) z . How is the RCS cooled during refueling operations with the refueling cavity full if BOTH RHR pumps fall to operate?
RESPONSE
We request that you accept as an additional answer that by maintaining 23 ft above the reactor vessel change an adequate heat sink la provided enough time to initiate emergency procedures to cool the core. REFERENCES - - - - - - - - - - - - - - - - - - - - - -- - - - - - - B.S.10.7
\
O l ! i l 5 i s., l
- R 1-5/dsp
~
02/06/86 Pgga 3 NRC Exam Section 7 ) QUESTION SRO 7.16: (SRO Requal 7.10)
)
Unit 3 is shutdown,4KV Bus 3A is deenergized, a EDG and #3 startup transformer are both INOPERABLE. It is required that certain vital loads on Bus 3A be operated. List three methods (including power source and any interim busses) by which this bus can be reenergized.
RESPONSE
We request that you add three additional methods by which the 3A bus may be energized to the answer key. _ _ _ _ _ _ _ _ _ _ _ _ _
- 1) B-EDG via the 3B I)us via A to B bus tie breaker
- 2) The switchyard via the main and aux. transformers. (Disconnect links can be outages.
g"ay; i (, 3) The switchyard via the units 1&2 startup transformer via .the cranking diesel bus
'"*' via the 3C 4KV bus to the 3A 4KV bus.
REFERENCE:
Drawing 5610-T-E-1591
,M K
- R I-5/dsp
.. ,. Q2/06/86 P:ge1 NRC Exam Sectiori 8
- QUESTION SRO 8.04: (SRO Requal 8.03) h,,
Using the attached technical specification, which action bekow would be correct for the
- following situation? SITUATION: Unit 3 at 50% power, Unit 4 in Startup mode with Tavg = 410 Deg, C AFW pump is taken out of service due to a surveillance (All other AFW equipment is OPERABLE)
- a. Unit 4 must be cooled down to < 350 degrees within 12 hours,
- b. EITHER Unit 3 OR Unit 4 must be shutdown / cooled down to < 350 degrees within 72 hours.
- c. BOTH Unit 3 and Unit 4 must be shutdown and cooled down to < 350 degrees within 12 hours if C AFW pump cannot be restored within 72 hours.
d.- Unit 3 must be shutdown and cooled down to < 350 degrees within 12 hours if C AFW pump cannot be restored within 72 hours. e._No action is required.- - - _ _ _ _ _ _ _ . _ - - _ _ _ _ _ -. _ __ _ _ _
RESPONSE
__ We feel that answers A or D should be accepted as correct. The reason for our request is stated below. The following discrepancies are noted in this question:
- 1. Not all the appilcable Technical Specifications necessary to evaluate this situation were included. Specifically section 3.0 was not included.
- 2. Unit 4 cannot be in the Startup Mode (Mode 2) when Tave = 4100 F. The min.
0 temp. for critically of Unit is 422 F, per OP-0202.2 pg. 5, paragraph 5.1. These discrepaneles led candidates to a variety of conclusions. Answer A would be the most correct under the following conditions:
- 1. Unit 4 is in Mode 2, Startup mode
- 2. Unit 3 is in Mode 1, Power operations
- 3. T.S.3.8.4 (b) is applicable because both units are > 3500F
- 4. T.S.3.8.5, action requirements, is not applicable to unit 4 because it is not in Mode 1,' Power Operation.
- 5. Therefore 3.0.1 is applicable requiring Unit 4 to be in at least Mode 3 within 6 hrs and to be in Mode 4 within the following 6 hrs thus requiring Unit 4 to be < 3500F within 12 hrs.
Answer D would be most correct if only the Tech Spec's attached to the exam were considered. This would require that section 3.8.5.b be considered valid for both units. 0 Thus one unit must be shutdown and cooled down to < 350 F within 12 hrs if the i inoperable AFW pump cannot be restored within 72 hrs. l
*RI-5/dsp
_ _ _ _ , -_. _ _.- .~ _.. _ ._ _ _ _ .
02/06/86 P:ga 2 NRC Exam Section 8 QUESTION SRO 8.13: (SRO Requal 8.09) D_, . List the three meteorological conditions which would preclude conducting a routine Gaseous Waste release. RESPONSE: , We request that you accept as a correct answer for wind direction "Toward the Girl ; Scout Camp", since this would require the wind to be from the south, and "Toward the
. Boy Scout Camp", since this would require the wind to be from the Northeut. .!
l
-See the attached site plan. - - - - - - - - - - - - ---- - - - - - - - - - - - - - - - - - >
l 1
REFERENCE:
l FSAR Turkey Point Plant Units 3 & 4 Vol.1 Fig. 2.2-3 0 l I I l l l l
- R1-5/dsp
l .- .- Q2/06/86 Pege 4 NRC Exam Section 8 QUESTION SRO 8.18: (SRO Requal 8.12) List the DNB related parameters as stated in Tech Specs, and their setpoints. (Assume normal power operations) l
RESPONSE
The NRC answer key has an apparent typographical error indicating the DNB limit on Pressurizer Pressure in units of psig instead of pela.
REFERENCE:
T.S.3.1.6.b l l
- RI-5/dsp
._ _ . ~ _
02/06/86 P ga 5 NRC Exam Section 8
- h. ' I QUESTION SRO 8.20: (SRO Requal 8.13) a) Where is the access code for the Autodialer retained?
b) How is operability of the Autedialer verified prior to being utilized in an Emergency situation?
, e- RESPONSR:
;,43 D
We request this question be deleted because no training was provided on the use of the Autodialer. Training was not conducted on this because the Autodialer has been and
~ - ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ - ~ ~ ~ ~ - -
remains inoperable.
REFERENCE:
FPL Letter Dated February 24,1983
- ' ~ig .
SI l
,.f[
t
,, r l
l l I
*RI-5/dsp
ENCLOSURE 5 REQUALIFICATION PROGRAM EVALUATION REPORT Facility: Turkey Point Nuclear Plant Examiner: William M. Dean, William G. Douglas Date(s) of Evaluation: February 3-11, 1986 Areas Evaluated: X Written X Oral N/A Simulator Written Examination
- 1. Overall evaluation RO:0 of 4 pass SRO: 5 of 9 pass
- 2. Evaluation of facility examination gradin:: N/A Oral Examination
- 1. Overall evaluation RO:3 of 4 Pass SRO:6 of 9 pass
- 2. Number conducted 13 Simulator Examination
- 1. Overall evaluation N/A
- 2. Number conducted Overall Program Evaluation Satisfactory Marginal Unsatisfactory X (List major Overall Pass Rate: 4 of 13 = 30.8?o deficiency areas with brief descriptive comments)
Submitted: Forwarded: Approved: f 3puh.Y.G g NilliamM. Dead 3pAAk Bruce A. Wilson Caudie A. Julian
~
Examiner Section Chief f Branch Chief A}}