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Exam Rept 50-250/OL-86-01 on 860203-12.Exam results:9 Reactor Operators & 13 Senior Reactor Operators Passed Oral & Written Exams.Total of 8 Reactor Operators & 11 Senior Reactor Operators Passed Complete Exam
	ML20203D236
Person / Time
Site:	Turkey Point
Issue date:	04/14/1986
From:	Bill Dean, Munro J; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML20203D223	List: ML20203D218; ML20203D236;
References
	50-250-OL-86-02, 50-250-OL-86-2, NUDOCS 8604220066
	Download: ML20203D236 (200)
	v • d • e

{{#Wiki_filter:- 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 Nuclear Plant Facility Docket No.: 50-250 Written examinations were administered at Redlands Country Club near Homestead, Florida. Oral examinations were administered at Turkey Point Nuclear Plant near Homestead, Florida. Chief Examiner: Otm,lm b d CC Y// y//G William M. Dean 'Date Signed 9'/o/K Approved by: [ 7 John F. Munro,' Acting Chief ~Date Signed Sumary: Examinations conducted February 3-12, 1986 Oral and written examinations were administered to twelve reactor operator (RO), fifteen Senior Reactor Operator (SRO) and one Instructor Certification (IC) candidates. Nine R0s and thirteen SR0s passed the written examination. Nine R0's and thirteen SR0s passed the oral examination. A total of eight R0's and eleven SR0's passsed the complete examination. 8604220066 860417 DR ADOCK 05000250 PDR L. _ _ _ _ _ _ _ _ - .-__s

REPORT DETAILS 1. Facility Employees Contacted: l

W. Miller, Training Supervisor (Nuclear)
C. Baker, Plant Manager

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B. Acosta, QA Supervisor
J. Crockford, Assistant Superintendent Nuclear Operations
P. Baum, Operations Training Supervisor
J. Shepard, Hot License Class Instructor
L. Goebel, Requalification Supervisor
Attended Exit Meeting 2.

Examiners:

W. M. Dean T. Rogers C. A. Casto T. Vinnola (EG&G)

F. Jagger (EG&G) B. Picker (EG&G) B. Heming (EG&G) l

Chief Examiner 3.

Examination Review Meeting At the conclusion of the written examinations, the examiners provided the l training staff with a copy of the written examination and answer key for review. The coments made by the facility reviewers are included as to this report, and the NRC resolutions to these comments are listed below, a. SR0 Exam (applicable R0 exam questions are in parenthesis) (1) Question 5.03 (1.03) NRC Resolution: Based on new reference material provided, recom-mended answer will be accepted. (2) Question 5.13 NRC Resolution: Will accept 12 steps as equivalent to 15 inches based on material in Technical Specification bases presented by the facility.

2 APR 171986 (3) Question 5.14 (1.14) NRC Resolution: Recommended answer will be accepted. (4) Question 5.18 NRC Resolution: Based on the different equation for axial offset presented in new training material provided to the NRC, recom-mended answer will be accepted. (5) Question 5.22 NRC Resolution:. Recommended additional answer will be accepted. (6) Question 5.23 (1.21) NRC Resolution: Answer key was based on a different facility's program temperature. Answer key will be modified to reflect program at Turkey Point. (7) Question 6.05 (3.03) NRC Resolution: Based on the fact that the subject instrument is not as yet operable and training has not been accomplished, question is deleted. (8) Question 6.10 (3.12) NRC Resolution: Based on new system drawings provided to the NRC, recommended additional answers will be accepted. (9) Question 6.11 NRC Resolution: Based on error in system descriptions, part 2 of this question is deleted. The answer to part 3 will be changed to upender. (10) Question 6.14 (3.17) NRC Resolution: Based on additional information presented in Technical Specifications that is not contained in the system descriptions, additional answer will be accepted. (11) Question 6.19 (2.20) NRC Resolution: Based on the minimal information presented to the candidates by the facility and the fact these modifications are not as yet installed, question is deleted. m . ~.

3 APR 171986 (12) Question 6.20(2.21) NRC Resolution: Based on recent changes to facility that were not in material provided to the NRC, the answer key will be modified as recommended. (13) Question 6.21 (2.22) NRC Resolution: Recommended additional answer will be accepted based on material provided. (14) Question 6.22 (3.20) NRC Resolution: Recommended additional answer is similar to existing answer key. No change is required. (15) Question 6.23 (3.21) NRC Resolution: Recommended additional answer is similar to existing answer key. No change is required. (16) Question 7.08 (4.09) NRC Resolution: Based on additional information provided to the NRC, recommended additional answers will be accepted. (17) Question 7.09 (4.12) NRC Resolution: Based on additional information provided to the NRC, only 2 of the 3 correct responses will be required for full credit. (18) Question 7.14: NRC Resolution: Based on the vagueness of the question and the naterial provided by the facility, the recommended additional answer will be accepted. (19) Question 7.15 NRC Resolution: Based on the new procedure that was not initially provided to the NRC, the answer key will be changed to reflect current guidance. (20) Question 7.16 NRC Resolution: Based on more detailed electrical system drawing provided to the NRC, recommended additional answers will be accepted.

4 APR 171986 (21) Question 7.20(4.18) NRC Resolution: Based on recent changes to a procedure that was not initially provided to the NRC, the recommended additional answer will also be accepted. (22) Question 7.21 (4.19) NRC Resolution: Recommended additional answer is similar to existing answer key. No change required. (23) Question 7.22(4.20) NRC Resolution: Recomended additional answer is similar to e xisting answer key. No change required. (24) Question 8.04 NRC Resolution: The subject Technical Specification is confusing in its application of " power operations" and " unit operations" and should be clarified. Technical Specification 3.0.1 should be familiar to SR0s. Mode 2 operation can be conducted at the given condition, though criticality can not be achieved before 522 F is reached. Literal interpretation of TS 3.8.5 leads to answer d, but due to inherent confusion in this TS, answer a will also be accepted as it is a feasible action which the operator may legitimately pursue. (25) Question 8.13 NRC Resolution: Recomended additional answer will be accepted based on area map provided by the facility. (26) Question 8.17 NRC Resolution: Based on additional information provided to the NRC supporting the non-applicability of this knowledge for SR0s, the question will be deleted. (27) Question 8.18 NRC Resolution: Typographical error noted will be corrected. (28) Question 8.20 NRC Resolution: Based on the continued non-operability of subject communications device and documentation supporting this fact, the question will be deleted.

5 (29) Question 8.23 NRC Resolution: Recommended additional answer will be accepted as it meets the intent of existing arswer key. b. R0 Exam (30) Question 1.13 NRC Resolution: Recommended additional answer is similar to existing answer key. No change required. (31) Question 1.16 NRC Resolution: Based on new material provided to the NRC, the recommended change to the answer key will be made. (32) Question 1.17 NRC Resolution: The test question asks about the SDM "calcu-lation" to which no correction is made. Answer key stands as is. (33) Question 1.22 NRC Resolution: Based on typographical error in the formula provided the candidates, question will be deleted. (34) Question 2.01 l NRC Resolution: Based on system drawing provided to the NRC, the recommended additional answer will be accepted. (35) Question 2.03 NRC Resolution: Revised logic drawing differs from that provided NRC in system descriptions. Change to answer key will be made. l (36) Question 2.08 NRC Resolution: Based on differences in the two Units shown on l the facility-provided drawing which is not reflected in the facility's system descriptions, recommended additional answer will be accepted. (37) Question 2.16 NRC Resolution: Based on logic diagram provided to the NRC which is not reflected in system descriptions, recommended changes to answer key will be made. L

1 6 APR 171986 (38) Question 2.17 NRC Resolution: Based on new material provided to the NRC, additional recommended answers will also be accepted. (39) Question 3.04 NRC Resolution: Based on new material provided which corrects error in system descriptions, answer key will be modified to reflect recomended change. (40) Question 3.08 NRC Resolution: Based on new material provided to the NRC, recommended additional answers will be accepted. (41) Question 3.18 NRC Resolution: Based on new material provided to the NRC, recommended additional answers will be accepted. (42) Question 4.06 NRC Resolution: Facility should coordinate training emphasis and existing procedures. Due to the disparity in the two, both recommended answers will be accepted. (43) Question 4.07 NRC Resolution: Based on information provided during the exam by the proctors, the additional recommended answer will be accepted. (44) Question 4.11 NRC Resolution: Based on information contained in a new procedure provided NRC, recommended additional answers will be accepted. (45) Question 4.13 NRC Resolution: Based on a new procedure not initially provided to the NRC, answer key will be modified to reflect recent change. c. Exam Questions Deleted During the Exam. In response to questions postulated by the candidates during the exam and subsequent investigation by the examiners, the following questions were deleted during the exam based on review of recent changes to facility that were not provided to the NRC: 6.03(2.04)and6.12(a). I

r 7 APR 171986 4. Exit Meeting At the conclusion of the site visit the examiners met with representatives F of the plant staff to discuss the results of the examination. Those individuals who clearly passed the oral examination were identified. There were generic weaknesses noted during the oral examinations. The areas of below normal performance were control board familiarity, knowledge of Rod i Control System, awareness of Emergency Response Guidelines background information and familiarity with various calculations applicable to plant performance (ECC, Heat Balance, SDM). i The cooperation given to the examiners and the effort to ensure an atmo-sphere in the control room conducive to oral examinations was also noted and appreciated. The licensee did not identify as proprietary any of the material provided to or reviewed by the examiners. i 4 i i i

matTg UNITED STATES <!m So NUCLEAR RE20LATORY COMMISSION 8' j REGION t1 6 g 101 MARIETTA STREET, N.W., SulTE 2000 ATLANTA, GEORGIA 30323 o. ~ p' ENCLOSURE 3 U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: TURKEY POINT 3&4 REACTOR TYPE: PWR-WEC3 DATE ADMINISTERED: 86/02/03 EXAMINER: DEAN, W H APPLICANT: INSTRUCTIONS TO APPLICANT: Uce separate paper for the answers. Write answers on one side oniv. Staple question sheet on top of the answer sheets. Points for each question are Indicated in parentheses after the question. The passins Stade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be pteked up sin (6) hours after the examination starts. i 0F CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY ___1_0___ _I'S 00_1__ ________ 5. THEORY OF NUCLEAR POWER PLANT 30 0 OPERATION, FLUIDS, AND THERMODYNAMICS ___1_0___ _['5 00 ________ 6. PLANT SYSTEMS DESIGN, CONTROL, 30 0 _1__ AND INSTRUMENTATION ___I_00____I'_5 00 ________ 7. PROCEDURES - NORMAL, A E:N O R M A L, 30 I__ EMERGENCY AND RADIOLOGICAL CONTROL _l5.00 ________ 8. ADMINISTRATIVE PROCEDURES, 30.00 CONDITIONS, AND LIMITATIONS 120.00 100.00 TOTALS FINAL GRADE _________________% All work done on this examination is my own. I have neither givcn nor received aid. 5EPL5C5U5I5~5E5U55UR5~~~~~~~~~~~~~~ d

Se negg gwgTED STATES O / o, NUCLEAR REGULATORY COMMISSION [ 'n REGION il 3 g 101 MARIETTA STREET, N.W SUITE 2000 %,......o*[ ATLANTA, GEORGIA 30323 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 2 QUESTION 5.01 (1.00) Which expression below describes the heat flux hot channel factor Fq(Z)? assemblybatheightZ/ Avg at height Z a. haximum fuel b. haximum fuel assembly 0 at height Z / Avg 0 in core c. Average fuel assembly b/ maximum batheight Z Averagebatheight Z/Avgbincore d. GUESTION 5.0; (1.00) Which of the following curves (see attached page) representing Xenon concentration is correct for the given power history? GUESTION 5.03 (1.00) 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 xenor, 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 m~a'g'nitude of the dip would be larger. QUESTION 5.04 (1.00) Which of the curves on the following page correctly represents the change in the critical boron concentration over core life? (***** CATEGORY 05 CONTINUED ON NEXT PAGE

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NUCLEAR RE20LATORY COMMIS$10N 8' nsosoM H i t os manseTTA Sineaf, m.w., Suitt 2000 $g.'.... c[ ATLANTA,osoncia seats ~ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3 QUESTION 5.05 (1.00) Which of the following would cause an inadvertant dilution accident? a) Overfilling a S/G while in hot standby. b) A Regenerative heat e:< c h a n g e r leak. c) Valving in a demineralizer that was not saturated. d) A VCT Lo-Lo level resulting in the RWST being used for charging. QUESTION 5.06 (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 operation. Which statement below correctly describes the effect on system volumetric flow rate and system head loss? a) Higher ftow rate, higher head loss b) Same flow rate, higher head loss c) Higher flow rate, same head loss d) Same flow rate, same head loss GUESTION 5.07 (2.50) The plant is operating at 30% power, turbine in AUTO (IhP IN), when lo'op 41 reactor coolant pump trips. Assuming a reactor trip does not occur, 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)

Tc in loop _91_, (0.5) 4) Th in loop 42 (0.5) 5) Nuclear Power (0,5) (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) f

>R Mhp#o, UNITED ST ATES NUCLEAR RET._ULATORY COMMISSION / E' ,, g j REoeoN il 5 g 101 MARIETTA STREET, N W., SulTE 2000 Aft.ANTA, GEORGIA 30323 s., -...../ ~ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND pac" GUESTION 5.08 (2.00) Indicate whether the following will cause the differential rod worth to INCREASE, DECREASE or have NO EFFECT. a) An adjacent rod is inserted to the same height b) Moderator temperature is INCREASED c) Boron concentration is DECREASED d) An adjacent burnable poison rod depletes GUESTION 5.09 (1.50) An ECC is caleviated for a startup 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. a) Xenon reactivity curve for trip from 6 0 *' is used to calevlate conditions to startup 20 hours after the trip. b) The Samarive reactivity curve is used instead of the :<enon r eactivity curve for startup 60 hours after trip. c) The power defect curve for 750 ppm is used instead of the 1450 ppm curve. QUESTION 5.10 (.50) Does the Latent Heat of Vaporization INCREASE, DECREASE or REMAIN THE SAME as saturation pressure / temperature of water is increased? (s*x** CATEGORY 05 CONTINUED ON NEXT P4GE

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f~ gm. 8800o, UNITED STATES f g [o, NUCLEAR RE ULATORY COMMISSION l o nEasoN il U g 101 MA#tETTA STnEET, N.W., SUITE 2000 ,8 ATLANTA. 0EoMGIA 30323 I 5. THEORY OF NUCLEKR POWER PLANT OPERATION, FLUIDS, AND PAGE 5 q QUESTION 5.11 (1.50) Unit 4 is at 50% power with control rods in hANUAL when the turbine is ramped up to 60%. Indicate whether the Parameters below will increase, decrease or remain the same during both the initial response (first 30 seconds of the transient) and after turbine power has stabilil:ed relative to the initial conditions. (Assume the following: No changes to boron / xenon Loop transport time 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) Teold d) Tavs QUESTION 5.12 (1.50) a) TRUE or FALSE: During cold plant conditions, you would enpect 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) GUESTION 5.13 (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? QUESTION 5 14 (1.00) List three significant heat transfer advantages of a counter flow heat oxchanger over a parallel flow heat exchanger. (***** CATEGORY 05 CONTINUED ON NEXT PAGE

- - - - )

e

a UNITED ST2TES /ja "Og,"o,, NUCLEAR REGULATORY COMMISSION E' MosoN il g 3 g 10140ARIETTA STMET.N.W SulTE 2900 f ATLANTA, OEORGIA 30323 o %, -......o* 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIOS, AND PAGE 6 QUESTION' 5.15 (2.00) Unit 3 has just restarted following a refueling outage while Unit 4 is naar EOL. Answer the following regardin3 the differences in plant response b9 tween the two units (e:: Plain your answers)* c) At a steady power level of 10EE(-8) amps during a startup, equal reactivity additions are made (approximately 100 pcm). 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 Tavg? GUESTION 5.16 (1.00) What are the two reasons for shifting the SI mode from cold les rocireviation to hot les recirculation approximately 24 hours after a LOCA? QUESTION 5.17 (1.00) Attached are curves for overall power defect and Doppler only power defect indicating both EOL and BOL values for a cycle 1 core. These show Doppler defect becomes less negative over core life. Why then does the overall power defect become more negative? QUESTION 5.18 (2.00) A rod drops and sticks at the core mid position from full power conditions with all rods out. A Reactor Trip does not occur. If this condition were to persist for an extended period of time (well beyond T/S limits), what will be the effect on the Excore Axial Offset of the Power Range NI for the quadrant in which the dropped / stuck rod occurs. Include a discussion of nenon effects and-a definition of axial offset. QUESTION 5.19 (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? (***** CATEGORY 05 CONTINUED ON NEXT PAGE

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j REGloN il 3

e t ot M ARIETT A STREET, N.W., SUITE 2900 ATLANTA, GEORGIA 30323 o s., *- / 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 7 QUESTION 5.20 (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 cnd what is the approximate change in count rate that is observed? QUESTION 5.21 (1.00) l Arrange the following types of radiation in order of penetrating power from LOW to HIGH: l 1. Beta 2. Gamma 3. Neutron 4. Alpha l GUESTION 5.22 (1.50) Identify the segments on the actual T-S Rankine Cvele on the following page for the processes below. (example answer: d) segment 5-6) a) Reheat from rio i s t u r e Separators b) Heat addition from feedwater heaters c) Heat removed by the HP turbine GUESTION 5.23 (1.00) Using the attached steam tables, what is the amount of primary subcooling et the core exit if the p r e s su r i::e r is at 2235 psis and Tavg is 575 degrees? (assume normal operating conditions) (xxxx* END OF CATEGORY 05 xxxxx)

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/gf" "8 3:n[o, NUCLEAR RE'ULATORY COMMISSION UNITED STATES 8' j REQlON H 3 tot MARIETT A STREET, N.W., SulTE 2000 8 ATLANT A, GEORGIA 30323 o s., - / ~ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 8 GUESTION 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 caleviated peak centerline temperature shall not e::c e e d 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 tnat the core remains amenable to cooling. QUESTION 6.02 (1.00) Which of the following statements correctly describes the RHR System lineup when HOT LEG RECIRCULATION is established? (assume both trains of RHR are available) a. Both trains are used to suppl,. hot les recirculation exclusively. b. One train is used to supply hot leg re. circulation and the other train is used to continue cold leg rectrevlation. c. Both trains supply both hot and cold leg recirc 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). (***** CATEGORY 06 CONTINUED ON NEXT PAGE ****m)

a at4 UNITED ST ATES /ja "o, NUCLEAR REEULATORY COMMISSION S' REGION il f, 3 g 101 MARIETTA STREET, N.W., SulTE 2000 o, [ ATLANT A, GEORGIA 30323 9g' p' 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 9 k G U E "., !! C !' 4.CC ,1.^C Which of the following describes the NORMAL BACKL*P source of water to the Fire Protection System if electrical power was lost to the Fire Protection System Pumps? a. The Backup Service Water pump takes a suction on the Raw Water Tanks and is lined up to the Fire Frotection pump suction. b. A spoolPiece is installed between the Fire Protection System and the Service Water Pump discharge. c. The Screen Wash Pump Discharse is connected to a nearby fire hydrant using a hose. d. The Elevated Storage Tank is lined up to provide gravity feed to the Fire Protection Sysgem by opening valve 794 (normally closed). QUESTION 6.04 (1.00) Which of the following correctiv describes the actions of the AFW Flow Controllers AFTER receiving an initiation signal s. Since the AFW flow control console Hand Indicating Controllers (HICs) are set to a predetermined flow rate, the control valve will RAPIDLY OFEN 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 {he 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 initiallys the valve will SLOWLY OPEN to tne 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 stabilizes, OPEN SLOWLY to the pre-set position. (xxxxx CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx) i 1 1 o

8 "% UNITED STATES 6 / Io, NUCLEAR REZULATORY COMMISSION E' j REoeoN il 3 e 101 MARIETT A STREET, N.W., SulTE 2000 a ATLANTA.GEOmotA 30323 o f j ~ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 10 b SOE4 TION 4.05 'i.00F Which statement below correctly describes operation of the GAMMA-METRICS Neutron Flux Monitor in samma 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 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 gamma flux as it does in much lower radiation fields. QUESTION 6.06 (1.00) Which of the following statements correctly states the purpose of the AMBER light associated with the spent fuel building area radiation monitor channel modules? a. If the indicated radiation levels exceed a setpoint between the normal level and the high setpoint, this light actuates. b. If the indicated radiation levels exceed a setpoint above the high setpoint, this light actuates. c. If the indicated radiation levels drop below a setpoint between 0 and normal levels (indicating a circuit failure), this light actuates. d. If the detector source solenoid is energized, exposing the detec_ tor; source for a channel check, this light actuates. e. This is the power available light and is always on as long as power to the channel module exists. (***** CATEGORY 06 CONTINUED ON NEXT PAGE **mmx) t

,aaffo UNITED STATES gf o,, NUCLEAR REZULATORY COMMISSION y o REG 40 Nil e 101 MARIETTA STREET, N.W., SUITE 2000 e ATLANTA, GEORGIA 30323 s/ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 11 QUESTION 6.07 (1.50) Answer the following questions regarding ESFAS TRUE or FALSE: a) 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 i 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 6.08 (1.00) c) What is the purpose of the silver coating on the Reactor Vessel flange 0-rings? (0.5) b) TRUE or FALSE: If the ' REACTOR VESSEL FLANGE LEAKOFF HIGH TEMP' alarm actuates, the isolation valve on the inner 0-ring leak off line will automatically close. (0.5) GUESTION 6.09 (1.25) What is the purpose of the following precautions associated with operation of the Reactor Coolant Pumps? a) Do not open il Seal Leakoff isolation valves until RCS pressure is 3reater than 100 psig. (0.5) b) Do not open 41 Seal Bypass valves until 41 Seal Leakoff valves are open with > 50 psid across #1 seal. (.75) (***** CATEGORY 06 CONTINUED ON NEXT PAGE **mxx) I

gaaCon"o, NUCLEAR I.EIULATORY COMMISSION UNITED STATES

' 'g REGION li 2 a 101 MARIETT A STREET, N.W., SUITE 2000 8 ATLANTA, GEORGIA 30323 o %..** / 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 12 QUESTION 6.10 (1.50) e) Give the location and the number of UV relays that must be energized 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)

.,5H(.5 GUESTION 6.11 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 lift, the will stop the hoist a few inches higher. The upper limit position switch is set so that the bottom no::le of the fuel assertbly will clear the __________. /1.50H(l.0C QUESTION 6.12 Indicate what automatic actions, if any, occur when high level alarms are received on the following process radiation monitors: Fortiuvioces dg N

)

15 (Fioni G u i. Air b) R-14 (Plant Vent Gaseous Activity) c) R-10 (S/G Blowdown Liquid Activity) QUESTION 6.13 (2.00) What are the 4 conditions which must be met for the overpressure mitigation system (OMS) stitiTs lights to be ON? QUESTION 6.14 (1.00) List the 4 sets of ECCS related valves required to mitigate a LOCA which have their control power breakers racked out during critical operations. (xxxxx CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx)

m 880g UNITED STATES -a o NUCLEAR RE1ULATORY COMMISSION [' REGION il 3 101 MARIETTA STREET. N.W., SulTE 2900 o, ATLANTA, GEORGIA 30323 %, '.... f, 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 13 GUESTION 6.15 (1.50) The Bus clearing relays set-up the permissive to use the startup 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? QUESTION 6.16 (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). GUESTION 6.17 (1.50) List the 6 essential safety related loads served by the Component Cooling Water System (redundant components like heat exchangers are one load). (1.'t h QUESTION 6.18

1.005 c)

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 deg F, what action is required to reduce the temperature? (Include the frequency of the action) (0.5) 'i.5fn h b 9titMFHOP d lo a) What indication does a control room operator have that a fire dam'per has actuated? (0.5) b) Explain in detail the design features which allow a fire damper to auto close when required. (1.0) GUESTION 6.20 (2.00) c) Describe the-evnback 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) (**xxx CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx) o e

n UNITED STATES /ja 88Og,% NUCLEAR RE20LATORY COMMISSION [' REGION il f, 5 a 101 MAA6ETTA STREET, N.W., SUITE 2000 e ATLANTA GEOAGIA 30323 %...* / 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUhENTATION PAGE 14 I QUESTION 6.21 (1.50) c) What are the 2 sources of Borated water available for the Spent Fuel Pool? (0.5) b) What 2 design features of the spent fuel racks ensure criticality does not occur in the Spent Fuel pool? (1.0) GUESTION 6.22 (1.75) a) What consequences could be expected in the Rod Control System's DC Hold Cabinet if 2 or more groups of rod drive mechanisms were placed on hold power (excluding Control Bank D rods)? Explain your reasoning. (1.0) b) Why is there both a 125 VDC and a 70 VDC power supply in the DC Hold Cabinet? (.75) GUESTION 6.23 (1.00) The Alternate Source Transfer Switches associated with the recently installed 120 VAC inverters have key locks to prevent 2 switches of the same channel being selected to ALTERNATE at the same time. What are the the purposes behind this administrative key control? QUESTION 6,24 (.75) Recently, 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 and 4's MSIVs. What is this design problem? (xxxxx END OF CATEGORY 06 xxxxx)

UNITED STATES 8+# "8 ? o NUCLEAR RElULATORY COMMISSION "g e REmoN n 3 e 101 MARIETTA STREET, N.W, SulTE 2000 2 ATLANTA, GEORGIA 30323

...+

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 15 ~~~~ d656L66 6dL C6UTR6L'~~~~~~~~~~~~~~~~~~~~~~~ ~ R QUESTION 7.01 (1.00) The Response Not Obtained for the first immediate action of E0P-FR-S.1 ' Response to Nuclear Power Generation / ATWS' is to manually trip the reactor. If the reactor will not trip, then* 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 situations -below requires initiation of Emergency Boration? a. Following a Rx Trip, the rod position indicators show TWO rods which are NOT fully inserted. b. Rod Bank D Low l i m i '. Alarm is actuated. c. An uncontrolled RCS Heatup following a reactor trip occurs. d. An unexplained decrease in reactor power occurs while at 70% rated power. QUESTION 7.03 (1.00) Which of the following reasons correctly describes the basis for allowing RCP restart in EOP-FR-C.1 ' Response to Inadequate Core Cooling'. a. Helps to mix the Sl 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 r,ot alleviate inadequate core cooling. (xxxxx CATEGORY 07 CONTINUED ON NEXT PAGE xxxxx)

I UNITED STJ.TES ga #8:u9'e, NUCLEAR REZULATORY COMMISSION 6 [' REGION 11 f, 3 g 101 MARIETTA STREET. N.W., SulTE 2900 4 ATLANTA, GEORGIA 30323 % *'...

pl 7.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 16 ~~~~~~~~~~~~~~~~~~~~~~~~ ~~~~R 656L6656dL'66UTRUL QUESTION 7.04 (1.00) E0P-ES-0.3, ' Natural Cirevlation Cooldown with Steam Voids with RVLMS', allows concurrent cooldown and depressurization of the RCS. Which state-cent 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-i:ing, 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 F.S-0.4 vice concurrently. QUESTION 7.05 (1.00) Which of the following malfunctions will result in both a low Tavs indication and a low delta T indication? a. Hot leg RTD failed high b. Hot leg RTD failed low c. Cold leg RTD failed high d. Cold leg RTD failed low QUESTION 7.06 (1.00) ~ Which of the fo15owing would cause the greatest biological damage to a man? a. 0.1 Rad of Fast Neutron. b. 1 Rem of Gamma. c. 10 Rem of Beta. d. 0.05 Rad of Alpha. (xxxxx CATEGORY 07 CONTINUED ON NEXT PAGE xxxxx)

UNITED STATES / g #88:0g"o, NUCLEAR REZULATORY COMMISSION P' ' 7, mEGeoN il 5 y 1oi maattTTA STREET. N.W., SulTE 2000 ATLANTA, GEOM 01A 30323 s/ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 17 ~~~~RI555L 55CIL E5sTR5L QUESTION 7.07 (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. QUESTION 7.08 (1.50) 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 cccurs and NO operator actions are taken to mitigate the dilution) GUESTION 7.09 (1.00) List the two conditions (including setpoints) which determine Adverse Containment conditions. QUESTION 7.10 (2.00) What are the Unit 3 Plant Supervisor's immediate actions if the word is passed 'Fie in the control room, shift personnel report to assigned control room evacuation stations' ? QUESTION 7.11 ---- (1.50) E0P-ECA-0.0, ' 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? (***** CATEGORY 07 CONTINUED ON NEXT PAGE_*****) l

ma stor UNITED STATES gf o, NUCLEAR RE'f ULATORY COMMISSION REGONil 3 o 3 g 101 MARIETTA STREET, N.W., SulTE 2000 8 ATLANTA. GEORGIA 30323 s.,... * / 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 18 ~~~~~~~~~~~~~~~~~~~~~~~~ ~~~~ A5i5t55iEAL 55ETR5L R QUESTION 7.12 (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 7.13 (1.50) List the three conditions stated in Tech Specs which make a control rod INOPERABLE. QUESTION 7.14 (1.00) How is the RCS cooled during refueling operations with the refueling cavity full, if BOTH RHR pumps fail to operate? GUESTION 7.15 (1.50) Besides required notifications, what are the immediate operator actions if you are on shift in the control room and the refueling supervisor in the containment reports they dropped a spent fuel element in containment? GUESTION 7.16 (1.50) Unit 3 is shutdown, 4KV Bus 3A is deenergized, A EDG and 43 Startup transformer are both INDPERABLE. It is required that certain vital loads on Bus 3A be operated. List three methods (including power source and any interim buses) by which this bus can bae reenergized. QUESTION 7.17 (1.00) Suberiticality and Core Cooling are the two highest priority CSF Status Trees to monitor _ during accident conditions. List the remaining 4 CSF Status Trees in OECREASING order of priority. (xxxxx CATEGORY 07 CONTINUED ON NEXT PAGE xxxxx)

united status

pa atog'o,,

NUCLEAR RE!ULATORY COMMISSION [' RE000N le g f, f a 101 MARIETTA STREET, N.W., SulTE 2000 ATLANTA, GEORGIA 30323 o s/ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 19 ~~~~ A5i5t55iCit 55sTR5t R QUESTION 7 18 (1.50) After Natural Circulation has been establishede what 3 indications are conitored to determine RCS COOLDOWN, according to ES-0.2, ' Natural Cireviation Cooldown'? OUESTION 7.19 (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) c) Verify AFW pumps running (0.6) GUESTION 7.20 (1.50) A failure in tne Unit 4 PZR level control circuitry results in level controllers LC 459C & 460C sending false Lo-Lo level signals (< 14%) to the PZR heater controllers. Describe what must be done to the Group A Backup Heaters in order to restore them to operation with the false level signal in place. Indicate locations of any controls operated and any indications in the control room that any remote controls are operated. QUESTION 7.21 (1.50) During a small break LOCA (SBLOCA), it is recotred to trip the RCP if the trip criteria are met. If forced flow throu3h the core promotes cooling, why are the RCPs tripped. _. ~ - - (***** CATECORY 07 CONTINUED ON NEXT PAGE

- - - - )

1 A

UNITED STATES km s tIg#'o NUCLEAR RESULATORY COMMISSION REGION 11 g 3 g 101 MARIETTA STREET, N.W., SulTE 2000 f ATLANT A, GEORGIA 30323 +., * - / 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCf AND PAGE 20 ~'~~E3656L66565[~66NTR6E'~~~~~~~~~~~~~~~~~~~~~~~ GUESTION 7.22 (2.00) Answer the following questions regarding EOP usage: a) 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 if, during performance of steps in a ORP (Optimal Recovery Procedure). an ORANGE terminus on a CSF Status Tree is encountered? (1.0) (xxxxx END OF CATEGORY 07

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~~

ha me 2g, UNITED ST.* TES / "o, NUCLEAR RECULATORY COMMISSION 8'

7, nEGeoM N 3

i iot man ETTA SinEET.N.W., SUITE 2900 8 ATLANT A, GEOnGIA 30323 s., / 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 21 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 QUESTION 8.02 (1.00) Which of the following =tatements correctly describes proper status of containment building penetrations during refueling operations? a. 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. b. Penetrations leading from the containment atmosphere to the outside atmospnere can be OPEN as long as an OPERABLE automatic isolation valve is in place. c. The equipment door is OPEN but capable of being immediately shut. d. The Containment Purge system is SECURED and all isolation valves SHUT. ~ (xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx) I f

e 884 UNITED STATES m q'o, NUCLEAR RE1ULATORY COMMISSION / F' 'k REGION il E 101 ht ARIETTA STREET, N.W.,5UITE 2900 f ATLANTA, GEORGIA 30323 o p' 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 22 GUESTION 8.03 (1.00) According to Tech Specs, which of the following is the correct action to be taken if the Radweste Effluent honitoring Line Process Monitor is out of cervice? a. Effluent releases cannot be performed until the Monitor is back in service. b. Effluent releases may be performed if Grab Samples are analy:ed every hour during the release. c. Effluent releases may be performed provided two samples taken prior to the release are analyzed and do not exceed 10CFR20 limits and two qualified staff members verify the release rate calculations and the discharge valve lineup. d. The effluent release may be performed provided a sample prior to the release indicates that the Lower Limit of Detection (LLD) is not exceeded for all the analyses required and subsequent grab samples during the release confirm this condition continues to exist. QUESTION 8.04 (1.00) Using the attached technical specification. which action below would be correct for the following situation? SITUATION: Unit 3 at 50% power, Unit 4 in Startup mode with Tavs = 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 s degrees within 72 hours. c. BOTH Unit 3 and Unit 4 must be shutdown and cooled down 350 degrees within 12 hours if C AFW pump cannot be restored wi thin-70 hour s, 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. (***** CATEGORY 08 CONTINUED ON NEXT PAGE

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' h.5~h 3.8 STEAM AND POWER CONVERSION SYSTEMS Apolicabilityt h!!as to the operating status of the steam and power conversion systems. Objective To define conditions of the steam-relleving capacity and auxillary feedwater system. Soecification: 1. When the reactor coolant of a nuclear imit is heated above 3500F, the following conditions must be met

a. TWELVE (12) of its steam generator safety valves shall be operable (except 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 or less.

d. System piping, interlocks and valves directly associated with the related components in TS 3.3.1 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 3 [~ specifications cannot be met within 48 hours, the reactor shall be ~ shutdown and the reactor coolant temperature redaced below 3500F. Specification 3.0.1 applies. 4. The following number of Independent steam generator auxillary feedwater trains and their associated flow paths (steam and water) sha!! be operable when the reactor coolant is heated above 3500F: (/ 7 h,O 'l o f b 104 3.8 1 Amendment Nos.110 and 1

a. Sinale Nuclear Unit Operation i

Two independent auxillary feedwater trains capable of being powered from an operable steam supply,

b. Dual Nuclear Unit Operation Two independent auxillary feedwater trains and a third pump capable of being powered from, and supplying water to either

~. traln. f ~

c. If in accordance with T5 4.10.1, testing is required during start-up

~ of either unit, TS 3.8.4.a. or b., as applicable, shall apply for an auxiliary feedwater pump, pumps, or associated flow oaths (steam and water) found to be Inoperable. 5. During power operation, if any of the conditions 'of 3.8.4 cannot be met, the reactor shall be shutdown and the reactor coolant temperature reduced below 3500F,.unless one of the following conditions can be met:

a. For single unit operation with one of the two required independent auxillary feedwater trains inoperable, restore the Inoperable train to operable status within 72 hours or the reactor shall be shutdown and the reactor coolant temperature reduced below 3500F within the next 12 hours,

b. For dual unit operation, one auxiliary feedwater pump and its associated piping, valves, and interlocks may be Inoperable provided two Independent auxillary feedwater trains remain operable for time period not to exceed 72 hours. If the Inoperable pump cannot be made operable within 72 hours, one reactor shall be shutdown and its reactor coolant temperature reduced below 3500F within the next 12 hours.

c. For dual unit operation, with one irxiependent auxillary feedwater i

train inoperable in one reactor, the affected reactor shall be 5HUTDOWN and its reactor coolant temperatu e reduced below 3500F within 72 hours. TS 3.3.5.a applies for the single unit still .e In operation. g I

d. For dualimit operation, with one independent auxillary feedwater train inoperable in both units, one reactor shall be SHUTDOWN l

and its reactor coolant temperature reduced below 3500F within l-12 hours. TS 3.8.5.a app!!es for the single unit stillin operation. 1 i l l 7 Of 1 i 3.8-? . Amendment Nos. 110 and 104

~ m 08:q, UNITED STATES a "o, NUCLEAR RE1ULATORY COMMISSION S' REG @N il o 3 g 101 MAmtETTA STREET.N.W., SulTE 2000 I ATLANTA, GEORGIA 30323 .o' ~ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 23 QUESTION 8.05 (1.00) According to Tech Specs, the secondary activity limit is _____ micro-curies per 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 GUESTION 8.06 (1.50) Answer TRUE or FALSE to the following. a. Entry into an Operational Mode may be performed even if the conditions for the Limiting Condition for Operation (LCO) are NOT met provided the ACTION requirements are subsequently satisfactorily completed. b. If a LCO is NOT met 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 e:: tension) shall constitute a failure of the component to meet its operability requirements. QUESTION 8.07 (1.50) Indicate whether the following situations violate the Technical Specification for Power operations on UNIT 4: a) 'A' EDG out of service with Unit 3 shutdown and 3A 4KV Bus deenergized. b)

B' EDG out 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.

(xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx) 1 f

jaa 8824q, UNITED STATES e ~'o,, NUCLEAR RE'_ULATORY COMMISSION 8' REGION 11 o 3 g 101 MARIETTA STREET. N.W., SUITE 2000 a ATLANT A, GEORGIA 30323 o s/ +...* 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 24 QUESTION 8.08 (1.50) Match the description of responsibility for administration of the tagging / clearance program in Column A with the appropriate individual in Column B. COLUMN A COLUMN B a) Denotes if Independent Verification is

1) RCO required.

2) PS-N

3) Maintenance SPVSR b)

Normally signs the Clearance Release part

4) NWE of the Equipment Clearance Order form.

5) Maintenance Worker c)

Enters / Deletes entries for Tech Spec related equipment in the Equipment Out of Service book. QUESTION 8.09 (1.00) Provide the minimum number of individuals required by tech specs for the following positions to operate both units at full power. a) Plant Supervisor-Nuclear (PS-N) b) SR0(s).(NOT including the plant supervisor (s)) c) ______ R0(s) d) ______ A0(s) o) ______ STA(s) GUESTION 8.10 (1.00) List the two individuals who may authorize equipment clearances. QUESTION 3.11 ___- (1.00) Under what two conditions may the Operator at the Controls (OATC) leave the Surveillance Area (without being relieved of his duties)? (***** CATEGORY 08 CONTINUED ON NEXT PAGE **xxx)

pa nUg UNITED STi.TES 6 "o, NUCLEAR RE2ULATORY COMMISSION 8'

'g REGION il 3

a 101 MARIETT A STREET. N.W., SulTE 200( ATLANTA, GEORGIA 30323 o s., - / 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 25 QUESTION 8.12 (1.75) a) What are the 4 Spray / Sprinkler systems which must be OPERABLE if the equipment they protect is required to be OPERABLE? (1.0) b) What actions are required and in what time frame must they be taken if one of these sprinkler systems was to be INOPERABLE? (.75) GUESTION 8.13 (1.50) List the three meteorological conditions which would preclude conducting a routine Gaseous Waste release. QUESTION 8.14 (1.50) List 5 acceptable methods by which Independent Verification of electrical breaker alignments may be accomplished. GUESTION S.15 (1.50) List the personnel who comprise the Fire Brigade (by position), assuming they are required to respond to a fire in a Radioligically Controlled Area. GUESTION 8.16 (1.50) What are the three responsibilities of the Emergency Coordinator which CANNOT be delegated. -OUEIT!O" ?.17 ,MF hpb List the four methods available for determining the release rate for Off-Site Dose calculat ions and indicate which one is the preferred method. .. ~. GUESTION 8 18 (1.50) List the DNB related parameters as stated in Tech Specs, and their setpoints. (Assume normal power operations) (***** CATEGORY 08 CONTINUED ON NEXT PAGE

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r [pm 88:'g*'o, UNITED STATES NUCLEAR RE10LATORY COMMISSION ', 'o p REGION il U g 101 M ARIETTA STREET, N.W., SUITE 2000 ATLANTA, GEORGIA 30323 s/ 8. ADMINISTR ATIVE FROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 7 OUESTION 8.19 (1.50) List the MINIMUM requirements for each of the necessary electrical systems listed below in order to conduct a reactor startup on UNIT 3, assuming Unit 4 is shutdown. c) Unit 3 and 4 4160 KV Buses b) Unit 3 480 VAC Load Centers l c) Battery Chargers b - 'UESUOM 9 2^ (1.00.1-l c) Where is the access code for the Autodialer retained? (0.4) b) How is operability of the Autodialer verified prior to being utilized in an Emergency situation? (0.6) QUESTION 8.21 (2.004 List 4 of the 5 general conditions in the facility's 'Re-Entry

emergency procedure under which the Emergenev C o o r d i ri a t o r may authorire entry into an area that has already been evacuated.

GUESTION S.22 (1.50) Technical Specification 3.4 allows all reactor coolant pumps and restdual heat removal pumps to be de-energized while in hot shutdown. a) What is the time frame that this condition may e:< i s t ? (0.5) b) What provisions must be met curing the period these pumps are de-energized? (1.0) QUESTION 8.23 (1.00) While performing the Power Range Nuclear Instrumentation Channel Functional Tost, a caution states that before taking a UNIT 4 power range channel out of servicer two conditions should be verified. List these two conditions. (**xxx END OF CATEGORY 08 xxxxx) (xxxxxxxxxxxxx END OF EXAMINATION xxxx*xxxxxxxxx*) I ~

Out)/(thergy in; o o og s o y,t - 1/2 at' I E o ac 2 KE = 1/2 av ,, gyf,.,,)fg 4, x; a, 3,-at 9 PE = men w = e/t x = an2/t1/2 = 0.693/tjjp yf = y, + at 1/2'N

M /' M 2

y,, 3p !! D A= [(t1/2)

II I)

S

E = 931 am m = V,,Ao

-n Q.=ph I*I'o Q = aCpat Q = UAa T I = I,e~"* Pwr = W ah !=I 10-x/WL f n TVL = 1.3/u P = P 10 ""III HVL = -0.693/u 8 p = p e/T t o SUR = 26.06/T SCR = S/(1 - K,ff) SUR = 26 {p + p)/(lie ff-p) CR, = S/(1 - K,ff,) SUR = 26e/t* + (s - o)T s CR (1 - K,ffj) = CR II ~ eff2) j 2 T = (1*/s) + [(a - oYIol M = 1/(I - Keff) = CR /CR, j T = s/(, - s) M = (1 - K,ff,)/(1 - K,ffj) l T = (s - o)/(Io) SDM = ( - K,ff)/K,ff a = (X,ff-1)/K,ff = 8Keff eff /K t= = 10 seconds I = 0.1 seconds-o = [(t*/(T K,ff)] + [a,ff (1 + IT)] / Ijj=Id d 2 =2 2 P = (rev)/(3 x 1010) Id Id jj g2 t = eN R/hr = (0.5 CE)/d'(meters) R/hr = 5 CE/d2 (f,,g) Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lem. 1 curie = 3.7 x 1010aps 1gaj.a3.78 liters 1 kg = 2.21 It,m 1 ft* = 7.48 gal. 1 np = 2.54 x 103 Btu /hr Density = 62.4 Itgi/ft3 1 nw = 3.41 x 106 5tu/hr Density = 1 ge/cW lin = 2.54 cm Heat of vaporization = 970 Btu /lem 'F = 9/5'C + 32 Heat of fusion = 144 Btu /lbe 'C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-lbf I ft. H = 0.4335 lbf/in. 2 e = 2.718 { e E '- y g I

l Volume, ft'/lb (Mhelpy. Stu.4b (Meepy,StuftbaF Typ To p ss. water tvep Steam t~eter tvep C2cm water

tve, Suom A

A A 8t s t as s eg a I - 's Fas 's g 32 0.08859 8.01602 3305 3305 -0.02 M75.5 1075.5 0.0000 2.1873 2.1873 32 35 0.09991 0.01602 2948 2948 3 00 1073.8 1076.8 0.006) 2.1706 2.1767 35 40 0.12163 0 0f602 2446 2446 8 03 1071.0 1079.0 0.0162 2.1432 2.1594 40 45 0.14744 0.01602 2037.7 2037.8 13.04 1068.1 1081.2 0 0262 2.1164 2.1426 45 80 0.17795 0.01602 1704.8 1704.8 18 05 1065.3 1083.4 0.0361 2.0901 2.1262 to 40 0.2561 0.01603 1207.6 1207.6 28.06 1059.7 1067.7 0.0535 2.0391 2.0946 60 70 0.3629 0.01605 868.3 868 4 38.05 1054 0 1092.1 0.0745 1.9900 2.0645 70 80 'O.5068

0.01607 633.3-633.3 48 04 1048 4 1006.4-

.0.0932. 1.9426. 2.0359 80 80 0.6981 0.01610 468.1 468.1 58 02 1042 7 1100.8 0 1115 1.8970 2.0086 60 300 0.9492 0.01613 350.4 350.4 68 00 1037.1 1105.1 0.129'i 1.8530 1.9825 100 110 1.2750 0.01617 265.4 265.4 77.98 1031.4 1109.3 0.1472 1.8105 1.9577 110 120 1.6927 0.01620 203.25 203.26 87.97 1025.6 1113.6 0.1646 1.7693 1.9339 120 130 2.2230 0.01615 157.32 157.33 97.96 1019.8 1117.8 0.1817 1.7295 1.9112 130 140 2.8892 0.01629 122.98 123 00 107.95 1014.0 1122.0 0.1985 1.6910 1.8895 240 150 3.718 0.01634 97.05 97.07 117.95 1008.2 1126.1 0.2150 1.6536 1.8686 150 160 4.741 0.01640 77.27 7729 127 96 1002.2 1130.2 0.2313 1.6174 1.8487 160 170 5.993 0.01645 62.04 62.06 137.97 996.2 1134.2 0.2473 1.5822 1.8295 170 180 7.511 0.01651 50.21 50.22 148.00 990.2 1138.2 0.2631 1.5480 1.8111 180 190 9.340 0.01657 40.94 40.96 158.04 984.1 1142.1 0.2787 1.514S 1.7934 100 200 11.526 0.01664 33.62 33.64 168.09 977.9 1146.0 0.2940 1.4824 1.7764 200 210 14.123 0.01671 27.80 27.82 178.15 971.6 1149.7 0.3091 1.4509 1.7600 210 212 14.696 0.01672 26.78 26.80 180.17 970.3 1150.5 0.3121 1.4447 1.7568 212 220 17.186 0.01678 23.13 23 15 188.23 965.2 1153 4 0.3241 1.4201 1.7442 220 230 20.779 0.01685 19.364 19.381 198.33 958.7 1157.1 0.3388 1.3902 1.7290 230 240 24.% S 0.01693 16.304 16.321 208.45 952.1 1160.6 0.3533 1.3609 1.7142 240 250 29.825 0.01701 13.802 13.819 218.59 945.4 1164.0 0.3677 1.3323 1.7000 250 l 260 35.427 0.01709 11.745 11.762 228.76 938 6 1167.4 0.3819 1.3043 1.6862 260 270 41.856 0 01718 10.042 10.060 238.95 931.7 1170.6 0.3960 1.2769 1.6729 270 290 49200 0.01726 8.627 8 644 249.17 924 6 1173.8 0.4098 1.2501 1.6599 280 290 57.550 0.01736 7.443 7.460 259.4 917.4 1176.8 0.4236 1.2238 1.6473 290 300 67.005 0.01745 6.448 6.466 269.7 910.0 1179.7 0.4372 1.1979 1.6351 300 310 77.67 0.01755 5.609 5.626 280.0 902.5 1182.5 0.4506 1.1726 1.6232 310 l 320 89.64 0.01766 4.896 4.914 290.4 894.8 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 1.5892 340 l 360 153.01 0.01311 2.939 2.957 332.3 862.1 1194.4 0.5161 1.0517 1.5678 360 1 i 380 195.73 0.01836 2.317 2.335 353.6 244.5 1195.0 0.5416 1.0057 1.5473 380 400 247.26 0.01864 1.8444 1.6630 375.1 825.9 1201.0 0.5667 0.9607 1.5274 400 420 30S 78 0.01894 1.4808 1.4997 396.9 806.2 1203.1 0.5915 0.9165 1.5060 420 440 381.54 0.01926 1.1976 1.2169 419.0 785.4 1204.4 06161 0.8729 1.4890 440 j 460 456.9 0.0196 0 9746 0.9942 441.5 763.2 1204.8 0.6405 0.8299 1.4704 460 410 566 2 0.0200 0.7972 0.8172 464.5 739.6 1204.1 06648 0.7871 1.4516 480 500 680.9 0.0204 0 6545 0.6749 487.9 714.3 1202.2 0.6890 C.7443 1.4333 500 520 812.5 _.._Q.0209 0.5386 0 55 % 512.0 687.0 1199.0 0.7133 0.7013 1.4146 520 540 962.8 0.0215 04437 0 4651 536 8 657.5 1194.3 0.7378 0.6577 1.3954 540 STO 1133.4 0.0221 0.3651 0.3871 562.4 625.3 1187.7 0.7625 0.6132 1.3757 560 580 1326.2 0.0228 0.2994 0.3222 589.1 589.9 1179.0 0.7876 0.5673 1.3550 580 400 1543.2 0.0236 0.2438 02575 617.1 550 6 1167.7 0.8134 0.5196 1.3330 GOO 820 1786.9 0.0247 0.1962 0.2208 646.9 506.3 1153.2 0.3803 0.46S9 1.3092 620 540 2059 9 0.0260 0.1543 0.1802 679.1 454 6 1133.7 0.8656 0.4134 12821 640 660 2365.7 00277 0.1166 0.1443 714.9 392.1 1107.0 0.8995 0.3502 1.2498 660 660 2708.6 0.0304 0.0808 01112 758 5 310.1 106P.5 0.9365 0.2720 1.2086 680 7G0 3094.3 0 0366 0.0386 0 0752 822.4' 172.7 9952 0.9901 0.1490 1.1390 700 705.5 32032 0.0508 0 0.0508 906.0 0 906.0 1.0612 0 1.0612 705.5 TABLE A.2 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (TEMPERATURE) s A.3 0

i Weaume. ft'/ib Enthelpy. Ste/tb titeepy. Der /e a F taergy,tv/te SS. I'*P geler Evep 94eem We'ter Evep Steam Water Evep Sleem Waerf Steem E'"'- Pole F pole

e A

s, s, s, e, 'e s messe 32.018 OA1602 3302.4 3302 4 0 00 1075.5 1075 5 0 2.1872 2.8872 8 5021.3 meses 3.10 35.023 SAlent 2945.5 2945 5 3 03 1073 8 10765 0 0061 2.1705 2.1766 SA3 10223 ane 0.15 45.453 8A1602 2004.7 2004 7 13 50 1067.9 1081 4 0 0271 21140 2 1411 13.50 1025 7 9.15 0 20 53.160 0 01603 1526 3 1526 3 21.22 10635 1084 7 0 0422 2 0778 2.1160 2122 102s3 gJe 0.30 64 484 0 01604 1039.7 1039 7 32.54 1057.1 1089 7 0 0641 2 0168 2.0009 32.54 1032 0 0.30 e 40 72.869 0.01606 792.0 792.1 40 92 1052.4 1093.3 0.0799 1.9762 2.0 W 2 40.92 1034.7 e.40 0.5 79.586 0 01607 641.5 641.5 47.62 1048 6 1096 3 0 0925 1.9446 2.0370 4742 1036 9 E5 0.6 85.218 0 01609 540 0 540.1 53 25 1045 5 1098 7 0.1028 1.9186 2.0215 5324 1038 7 0.6 u M.., - up, agages 0.00416% 466r.9b.,.48Et6 - 4 Seh6.1042 A 230 % -tLa4 <.1J066;3 6 1,A10ngsaa,3 < 4 ta, - 0.8 94.38 0.01611 411.67 411.69 62.39 1040 3 1102.6 0.1117 13775 1.9970 6229 1041.7 9.8 s.9 98.24 0 01612 368 41 36843 66 24 1038 1 1104 3 0 1264 1A604 1.9870 4624 1042.9 9.9 l 1.0 101.74 0.01614 333.59 333 60 69.73 1036.1 11058 0.1326 13455 1.9781 69.73 1044.1 1A 2.0 126 07 0.01623 173.74 173.76 94.03 1022.1 1116 2 0.1750 1.7450 1.9200 94A3 1051A 2A 3.0 14147 0.01630 118 71 118 73 109 42 10132 1122 6 0.2009 14854 1.8864 10941 1056.7 8.0 4.0 152 96 0.01636 90 63 90 64 120.92 1006 4 1127.3 0.2199 14428 1A626 120.93 1060.2 4.0 5.0 162 24 0.01641 73.515 73.53 130 20 1000.9 1131.1 0.2349 1.6094 1A443 130.18 1063.1 E.0 6.0 170 05 0 01645 61.967 61.98 138 03 996.2 1134.2 0 2474 1.5820 1A294 138.01 1065.4' 4.0 7A 176 84 0.01649 53 634 53.65 144 83 992.1 1136 9 0 2581 1.5587 1A168 144.81 1067.4 7A 8.0 182 86 0.01653 47.328 47.35 150 87 988.5 1139.3 0 2676 1.5384 1A060 15034 1069.2 a.0 9.0 18827 0 01656 42.385 42 40 156.30 965.1 1141.4 0 2760 1.5234 1.7964 156J8 10708 9.0 to 193.2) 0.01659 38 404 3642 161.26 982.1 1143.3 02836 1.5043 1.7879 16123 10723 10 14.696 212.00 0 01672 26 782 2683 183 17 970.3 1150.5 0 3121 1.4447 1.7568 180.12 1077.6 14.s96 15 213 03 0.01673 26.274 26.29 181.21 969.7 1150.9 0 3137 1.4415 1.7552 181.16 1077.9 15 20 227.96 0.01683 20 070 20 067 196 27 9601 1156 3 0.3358 1.3962 1.7320 19621 3082A to 30 250 34 0 01701 13 7266 13 744 218 9 945.2 1164.1 0 3652 1.3313 1.0995 215 3 1087.9 30 40 267.25 0 01715 10 4794 10 497 236.1. 933 6 1169A 0.3921 1.2844 1.6765 236 0 1092.1 40 50 26102 0.01727 84967 5514 250.2

923 9 1174.1 0 4112 12474 J.6586 250.1 10953 to 60 292 71 0.01738 7.1562 7.174 262.2 915 4 1177.6 0.4273 1.2167 1A440 242.0 1098.0 to 70 302 93 0.01748 6 1875 6 205 272.7 907A 1180 6 0 4411 1.1905 1A316 272.5 1100.2 70 80 312 04 0 01757 5 4536 5 471 232.1- " 900 9 1183 1 0 4534 1.1675 14208 281.9 1102.1 30 to 320 29 0 01766 4.8777 4.895 293 7 894 6 1185.3 0 4643 1.1470 1.6113 290.4 1103.7 90 100 327.82 0 01774 4.4133 4.431 258.5 888.6 1187.2 04743 1.1284 1.6027 298 2 1105.2 100 120 341 27 0 01789 3 7097 3 728 312 6 877.8 1193 4 0 4919 1.0960 1.5879 3122 1107.6 120 140 353 04 001803 3 2010 3 219 3250 868 0 1193 0 0 5071 1 0681 15752 3245 1109.6 140 360 363 55 0 0;815 2.6155 2 834 336.1 859 0 11951 0.5206 1.0435 1.5641 335.5 1111.2 160 180 373 08 0 01827 2.5129 2.531 346 2 850 7 1196.9 0 5328 1 0215 1.5543 3454.1112.5 180 200 351 80 0 01839 2.2689 2.287 355.5 842.8 1198.3 0 5438 1.0016 1.5454 3543 1113.7 300 250 400 97 0 01665 1.8245 1A432 376I 825 0 1201.1 0.5679 0 9585 1.5264 3752 1115.8 250 300 417 3b 0 018E9 1.5233 1.5427 394 0 8069 1202 9 0.5682 0 9223 1.5105 392.9 1117.2 300 350 431.73 0 01913 1.3064 1.3255 409 8 794 2 1204 0 0 60i! 08909 1.496'l 406 6 3118 !

350 400 444 60 00193 1.14162 1.1610 424.2 760 4 1204 6 0 6217 0 8630 1.4841 422.7 111E 7 400 450 4t4 28 0 0195 1.01224 .l.0318 437.3 767.5 1204.2 0 6360 08378 1.4738 435 7 1118 9 450 500 4E7 01 00199 0 90787 0 9276 4495 755.1 1204 7 0 6490 0 8145 1.4639 447.7 1118 8 900 553 476 93 00199 0 82183 0 8418 460 9 743.3 1204 3 06611 07936 1.4547 45E.9 1118 6 550 400 48523 0 0201 074962 07698 4713 732.0 12037 0E723 0 7738 1.4461 469.5 1116.2 500 703 .503 08 00205 063505 0 6556 4916 710 2 1201 8 0692R 07377 1.4304 488 9 1116 9 700 833 514 21 0 0209 0.54309 05690 509 8 689 6 1199 4 0 7111 0 7051 1.4163 5067 I115.2 000 900 ti! 95 0 02i2 0 4756B 05009 526 7 669 7 11 % 4 0 7279 06753 1.4032 5232 1113.0 900 3000 544.EB 0.0216 0 42435 0 4463 542 6 (50 4 1192.9 07434 06476 1.3910 SE6 1110.4 1000 1100 55C 2d 0.0720 0 37863 0 4005 557.5 631.5 11891 0 7573 0 6216 1.3794 5131 1107.5 1100 1200 l 57.19 0 0223 0 34013 0.3625 571.9 6130 1184 8 07714 05969 1.3683 566 9 1104.3 3200 1300 L7742 0 0227 0 30722 0.3299 585 6 594 6 1 ISO 2 0.7843 05733 1.3577 580.1 1100 9 1300 14c0 537 07 0 0?31 0 27871 0 3018 598 8 576 5 1175 3 07966 05507 1.3474 592 9 1097.1 1400 1500 5 % 20 0 0235 025372 0 2772 611.7 550 4 1170 1 0 8035 0 f 253 1.3373 605 2 1093.1 1500 2000 635 80 0.02's7

01676t, 01883 6721 4662 113B 3 08625 0 4256 1.7881 662 6 10GS6 2000 2500 65d 11 0 02c.f 010209 01307 731 7 3616 1093 3 C 9139 0 3206 12345 718.5 1032.9 2500 3000 695 33 0 0343 0 050/3 0 0850 801 8 218 4 1070 3 0 9728 01891 1.1619 7822 973.1 3000 3298 2 70147 00508 0

0 050d 906 0 0 906 0 10612 0 1.0612 875.9 875.9 3P081 TABLE A.3 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (PRESSURE) 1 A.4

Abe peese. Tea,seelwe, F ) 100 200 300 400 000 000 700 000 900 1000 1100 3200 1300 8400 3500 e 0.0161 502 5 4523 611.9 671.5 631.1 400 7 3 A es 00 lite 2 1195.7 1241.8 1208 6 1336 1 1984 5 (101.74) s 0.1295 2Ah09 2.1152 I1722 2.2237 2.2708 IJ144 00161 '7814 9024 102.24 114.21 12615 138 08 150 01 161 94 17346 185 78 197.70 209 62 221.53 233 45 6 a es 01 3548 6 1144 8 1241.3 1288 2 1335 9 the3 1433 6 1483 7 1534 7 1586 7 1639 6 1693 3 1744.0 1803 5 (167 24) s 0.1795 1A716 1.9369 1.9943 2.0460 20932 flE9 2 1776 2 2159 2 2521 2.76 % 2.3194 2 3509 2.3all 2A101 e 0 016) 38 84 44 93 51 03 57.04 63 03 69 00 74 98 80 94 86 91 92 37 98 34 los 30 110 76 116 72 to a 68 02 1146 6 11937 1240 6 12u7.8 1335 5 1384 0 1433 4 1483 5 1534 6 1546 6 1639 5 1693.3 1747.9 1803 4 (19.'.21) s 0.1295 1.7928 1.8593 1.9173 1.9692 2 0166 2.0603 2.1011 2 1394 2 1757 2.2101 2 2430 2.2744 2.3046 2.3337 i e 00161 0 0166 29 899 33 963 37.985 41 966 45978 49 964 53 946 57.926 61905 65 882 69358 73 833 77207' 16 6 68 04 164 09 1192 5 1239 9 1287.3 1335 2 13838 1433 2 1443 4 15345 15405 1639 4 19937 1747A 1803 4 (213.03) s 0 1295 0.2940 1.8134 18720 1.9242 1.9717 2.0155 2.0 % 3 2.0946 2.1309 2.1653 2.1982 2A297 2.2599 2.2390 ) e 0 0161 0 0lM 22356 25 428 28 457 31466 34 465 37.458 40 447 43 435 46 420 49 405 52J08 55.370 58.352 20 6 48.05 168 11 1191.4 1239 2 1286 9 1334.9 1383 5 1432 9 1483 2 1534.3 1986 3 1639.3 1803.1 17472 18033 (227.96) s 01295 0.2940 1.7805 1A397 13921 1.9397 1.9036 2 0244 2 0628 2.0991 2.1336 2.1665 2.1979 2.2282 22572 e 0 0161 0 0166 11 035 12 624 14.165 15 685 17.195 18 699 20 199 21 497 23 194 24689 26 183 27.676 29.148 40 6 48 10 168 15 1156 6 1236 4 1285 0 1333 6 1382.5 1432 1 1482.5 1533.7 15458 16388 1992 7 1747.5 1803 0 (267.25) s 0.1295 0 2940 1.6992 1.7608 1A143 1A624 1.9065 1.9476 1.9660 2.0224 2.0%9 2.0899 2.1224 2.1516 2.1807 e 0.0161 0 0156 7.257 8 354 9 400 10 425 1143S 12 446 13.450 14 452 15.452 16.450 17A4s 18.445 19.441 60 4 68 15 16520 1181 6 1233 5 1283 2 1332.3 1381.5 1431 3 1481.8 1533 2 15853 1638 4 1692 4 1747.1 1802A (292.71) s 0.1295 0.2939 1.6492 1.7134 1.7681 1A168 13612 1.9024 1.9410 1.9774 2.0120 2.0450 2.0765 2.1058 2.1359 e 0.0l61 0 0166 0 0175 6 218 7.018 7.794 & 560 9 319 10 075 10 829 11 581 12331 13ABI 13A29 14.577 30 6 68 21 168 24 269 74 1230 5 1281 3 1330.9 1380.5 1430 5 1481.1 1532 6 1584.9 1638 0 let2A 1746A 1802.5 (312.04) s 0.1295 0 2939 0 4371 1.6790 1.7349 1.7642 12289 1 8702 1.9089 1.9454 1.9800 2.0131 2A446 2.0750 2.1041 e 0 0161 0 0166 0 0175 4 915 5 588 6.216 6 833 7.443 8050 8 455 9.258 9 360 10A60 11AGO 11.459 100 6 68 26 168 29 269 77 1227.4 1279 3 1329 6 41379 5 1429 7 1480 4 1532.0 1544 4 1637.6 1991.6 1746.5 1902.2 (327.82) s 0.1295 0 2939 04371 1.6516 1.7083 1.7586 15036 13451 1.8839 1.9205 1.9552 1.9883 2A199 2.0502 2.0794 e 0 0161 0 01 % 0 0175 4 0786 4 6341 5.1637 5 6831 6.1925 6 7006 7.2060 7.7096 S.2119 8.7130 9.2134 9.7130 320 A 68 31 168 33 269 81 1224.1 1277.4 13281 1378 4 1428 8 1479 8 1531.4 1543 9 1637.1 1601J 1746 2 1802A (341.27) s 0 1295 0 2939 0 4371 1.6286 1.6872 1.7376 1.7829 13246 1 A635 1.9001 1.9349 1.9600 1.9996 2.f35JO 2.0592 e 0 0161 0 0166 0 0175 3 4651 3 9526 4 4119 4.8585 5.2995 5.7364 6 1709 6 6036 72349 7A652 7A946 S.3233 140 4 68 37 168 38 26985 1220 8 1275 3 1326 8 1377.4 1428 0 1479.1 1530 8 1583 4 1636 7 1890.9 1745.9 1301.7 (353 04) s 01295 02939 0 4370 1 6095 1.6686 1.71 % 1.7652 1.8071 13461 13828 1.9176 1.9508 1.9825 2.0129 2.0421 e 0 0161 0 0165 0 0175 3 0060 3 4413 3 8480 4.2420 4 6295 5 0132 5.3945 5.7741 6 1522 6 5293 ~6.9055 7.2011 ISO & 68 42 168 42 269 89 1217.4 1273 3 1325 4 1376 4 1427.2 1478 4 1530.3 1542.9 1636.3 190.5 1745.6 1801.4 (363 55) s 0 1234 02938 0 4370 1.5906 1.6522 1.7039 1.7499 1.7919 1A310 1A678 1.9027 1.9359 1.9676 1.9900 2.0273 e 0 0161 0 0166 0 0174 2 6474 3 0433 3 4093 3 7621 4.1064 4.4505 4.7907 5.1289 5 4657 52014 6.1363 6.4704 180 4 68 47 166 47 26S9/ 1213 8 1271.2 1324 0 1375.3 1426 3 1477.7 1529 7 1582 4 1635 9 1940 2 17453 1801.2 (373. Cal e C.1294 0.2538 04370 1 5743 1A376 3.6900 1 7362 1.7784 1A176 1 8i45 12894 1.9227 1 9545 1.9849 2.0142 e 0 0161 0 0166 0 0174 23598 2 7247 3.0583 3 3783 3 6915 4 0006 4.3077 4.612R 4.9165 52191 55209 5A219 200 A 68 52 ILS 51 269 96 1210 1 12690 13226 1374 3 1425.5 1477.0 15291 1581.9 1635 4 1889 8 17450 1800 9 (331.80) s 0 1294 C 293S 0 4359 1.5593 1.6242 1.677G 1.7239 1.7663 13057 1.8426 1A776 1.9109 1.9427 1.9732 2A025 e OC161 ~0 01C6 0 0174 0 0186 2.150: 24662 2 6872 2.9410 3 1909 3 4382 3 6837 39278 4.1799 4 4131 4.6546 250 6 68 66 168 03 270 05 3/5 10 1263 5 13190 1371 6 1423 4 1475 3 1527.6 1580 6 1634 4 1688 9 1744 2 10002 (400 97) s C 1294 02E7 04355 05567 1.5951 1.6502 1.6476 1.7405 1.7601 1 8173 1 8524 IAd58 1.9177 1.b482 1.9776 e t 0 0161 0 01E5 0 0174 0 0186 1.7666 2.0044 2 2263 2.4407 2.6509 2 8585 3 0643 3.2688 3 4721 3 4746 3376e 300 A I 68 79 1 % 74 27u 14 375.15 1257 7 1315 2 1368 9 1421.3 1473 E 152t 2 15794 1633 3 16880 1743 4 17993 (417.35) s 0 1294 0 2937 0 4337 C5%5 1.5703 1.6274 1.6758 1.7192 1.7591 3.7964 1A317 1A652 12972 1.9278 1A572 e 00161 0 0106 0 0174 0 018G 1.4913 1.7028 12970 2 0332 2 2652 2 4445 2.6219 2.7980 2.9730 3.1471 3J205 350 A 68 92 3(3 85 270 74 375 21 1251 5 1311 4 1366 2 1419 2 1471 8 1524 7 1578.2 1632.3 1647.1 17426 17989 (431 73) o 0 1293 0 29M 0 43G7 0 568,4 1.5483 1 6077 1.6571 1.7009 1.7411 1.7787 1 8141 1A477 13795 1.9:05 1.9400 e 0 0161 0 0166 0 0174 0 0162 1 2841 14763 1 6499 1.8151 1 9759 2 1339 2 2901 2.4450 2.5987 2.7515 2.9037 400 a 69 05 168 97 270 33 375 27 12451 1307.4 13634 14170 14701 1523 3 1576 9 1631.2 1686 2 1741 9 179S2 (444.60) s 51293 02935 0 4365 0 56G3 1.5282 1.5901, 1 6406 1.6850 1.7255 1.7632 1.7968 1.8325 12647 1.8955 1.9250 e 0 0161 0 0106 0 0174 0 0186 0 9919 1.1584 1 3037 1.4397 1.5708 1 6992 1 8256 1.9507 2.0746 2 1977 2.3200 500 A 69 32 159 to 270 51 3?S 38 1231 2 12991 1357.7 14127 1866 6 1520 3 1574 4 16291 1684 4 1740 3 1796.9 (457.01) s 01292 02334 04354 05t60 14921 1 5595 1 6'23 1 65/8 1 6990 1 7371 1.7730 1.8069 18393 13702 13998 TABLE A.4 PROPERTIES OF SUPERHEATED STEAM AND COMPRESSED WATER (TEMPERATURE AND PRESSURE) A.5

m ge m. h mpetInc,F R@sg in. pt.tsep) 100 300 300 400 600 400 700 000 900 3000 1100 3300 1300 1400 3500 e 00161 0 01 % 0 0174 0 0186 0 7944 0 94 % I M26 1.1892 1.3008 14093 1.5140 l ull I.7252 18264 1.9309 See a 69.58 th e2 270 70 375 49 12:59 3290 3 1951 8 3408 3 1463 0 1517 4 1578.9 1627.0 1682 6 1738 3 1795,.6 (48620) s 0.1292 82933 04M2 0 % 57 14590 1.5329 15844 16351 16769 1.7355 1.7517 1.7859 33384 13494 137 y e 0 0161 4 0166 _0 0174 0 0186 0 0704 0 7928 0 9072 3.0102 1.1078 12023 12948 3.3858 3.4757 1 % 47 1.6530 780 6 6934 let 65 270 89 375 61 447 93 1781 0 1345 6 1403 7 1459 4 1514 4 1%94 1624 8 leto 7 1737 2 1794.3 $03.C8) s 0 1291 82932 0 4360 0 % 55 0 6889 1.5090 1.5673 1 6154 14540 16970 1 7335 1 7679 15003 18318 18617 e 0 0161 0 0166 0 0174 0 0186 0 0704 0 6774 0 7825 0 8759 0 9631 1 0470 1.1289 1.2093 12875 13R9 1.4446 388 6 70.11 169 88 271.07 375 73 487Ad 1271.1 1339 2 13991 1455 R 1511 4 1%69 IM27 1678 9 !?36 0 1792.9 $182.). 0.1290 0 2930 0 4358 0 % 52 0 6885 1 4869 1 5484

1. % 80 1 6413 16807 1.7175 17522 3.78b1 18164 1 8464 e

00161 0 01G6 0 0174 0 0186 0 C2'J4 0 5869 0 6858 07713 0 8504 0 9262 09998 10720 11430 12131 1.2825 900 6 70 37 17010 271.2G 375 24 487 83 1260 6 1332 7 1394 4 1452 2 15065 1%44 IWO6 1677 1 17341 1791.6 l @ 31.95) s 0 1290 0.2929 04357 0 % 49 06881 1.4659 1.5311 1.5822 16263 1 M62 1.7033 3.7382 1.7783 18028 13329 i e 0 0161 0 01M 00174 0 0186 0 0204 05137 06060 0 6875 0 7603 0 8295 0 89 % 0 9622 1.0266 1.0901 1.1529 3800 6 70.63 170 33 273.44 375 96 44L79 1249 3 13259 1389 6 1444 5 1504 4 1561.9 16184 1675.3 1732.5 1790 3 (544.58) s 0.1269 02928 04355 0 % 47 0 6876 1.4457 1.5149 1.M77 1.6126 16530 14905 1.7256 1.7589 1.7905 1.8207 e 00161 0 01M 00174 00185 0 0203 04531 0 5440 0 6188 C 6865 0 7505 0 8121 0 8723 0 9313 0 9894 1.0468 i 1180 6 70 90 170.56 271 63 376 08 447.75 1237 3 1318 8 1384 7 1444 7 1502 4 1559 4 1616 3 1673.5 1731 C 1789 0 OS6.28) s 0.1269 02927 04353 0 5644 0 6872 1.4259 1.4996 1.5541 16000 14410 1.6787 1.7141 1.7475 1.7793 13097 e 0 0161 0 0166 0.0174 0 0185 0 0203 0 4016 0 4905 0 % 35 0 6250 0 6845 0 7418 07974 0 8519 0.9055 0 9584 3200 6 71.16 170 7A 271A2 376.20 487.72 1224 2 1311 5 1379 7 1440 9 1449 4 15 % 9 3614 2 16716 1729 4 1787.6 l @67.19)s 0.1288 02926 0.4351 0.5642 0.6868 1.4061 1.4851 1.5435 1.5843 16298 1 M79 1.7035 1.7371 1.7691 1.7996 ~ e 001(I 0 0166 0 0174 0 0185 0 0203 0.3176 0 4059 04712 0 5282 0 5809 0 6311 0 6794 0 7272 0.7737 08195 1400 6 71 68 17124 272 19 376 44 487 65 1194.1 12961 1369 3 1433 2 1493 2 15518 1609 9 1668 0 1726 3 1785 0 @ 87.07) s 0.1287 02923 0 4348 0.5636 0 6859 1J652 1.4575 1.5182 1.5670 140M 1.6484 1.6845 1.7185 1.7508 1.7815 l e 0.0161 0 0166 0.0173 0 0185 0 0202 0 0236 0.3415 0 4032 0 4555 0.5031 0 5482 0 5915 0 6336 0.6748 0.7153 1600 6 72 21 17169 272.57 376 69 487.60 616 77 J2794 1358 5 1425 2 148E 9 1546 6 1605 6 1664 3 1723.2 1782.3 (6C4 87) s 01286 0 2921 0 4344 0 5631 0 68bt 0.8129 14312 1 4963 1.5478 1.5916 1 6312 1.6678 1.7022 1.7344 1.7657 e 0 0160 0 0165 0 0173 0.0185 0 0202 0 0235 0 2906 03500 0 3988 0 4426 0 4836 0 5229 0 5609 0.5980 0 6743 1800 a 72.73 172.15 272.95 376 93 487.% 615.58 1261.1 13472 1417.1 1480 6 1541.1 1601.2 1660.7 1720.1 1779.7 (621/32) s 0.1244 02918 04341 0 5626 0.68'3 0 8109 1.4054 1.4768 1 5302 1.5753 1.61 % 14528 1 6876 1.7204 1.7516 e 0 0160 0 0165 0 0173 0 0184 0 0201 0 0233 0 2488 0 3072 0 3534 0 3942 0 4320 0 4680 0.5027 0.5M5 0.5695 2000 6 73 26 172 60 273 32 377.19 487 53 614 48 1240 9 1353 4 1408 7 1447.1 1536 2 1596.9 1657.0 1717.0 1777.1 (635 80) s 01263 02916 0 4337 0 5621 0 6834 0.8091 1.3794 14578 1.5138 1.5603 1.6014 1.6391 1.6743 1.7075 1.7389 e 00160 C.0165 0 0173 0 0184 0.0200 0 0230 0 1641 0.2293 02712 0.3068 03390 0 M92 03980 0 4259 0.4529 8500 6 74 57 173 74 274 27 377 82 487.50 612.08 1176 7 1303 4 1386.7 1457 5 1522.9 15859 16472 17092 1770 4 (668.11) s 0.1280 0 2910 0 4329 0 5609 0 6815 0 8048 1.3076 1.4129 1.47M 1.5269 1.5703 1.6094 144% 16796 1.7136 e 0 0160 0 0165 0 0172 0 0183 0 0200 0.0228 0 0982 0 1755 0.2161 0.2484 0.2770 0.3033 0.3282 0.3522 0.3753 3000 6 75 E3 17t SE 275.22 378 47 487.52 610 06 1060 5 1267.0 13632 1440.2 15014 1574A 1635 5 1701.4 17(1.8 (E95.33) s 0.1277 029.4 0 4320 0 5597 0 6796 0 8009 1.1966 1.3692 1.4429 1.4976 1.5434 1.5841 1 421.' 3.0561 14888 e 0 0160 0 0165 0 0172 0 0183 0.0199 0.0227 0.0335 0 1588 0 1987 0.2301 0 257t> 0.2827 0J065 0.3291 0.3510 3200 6 76 4 175.3 275 6 3787 487.5 609 4 800 8 1250 9 1353 4 1433.1 15038 1570 3 1634A 1698.3 1761.2 f.705.08) a C3276 0 2902 0.4317 0 5592 06768 0.7994 0 9708 1.3515 1.4300 1.48 % 1.5335 1.5/49 1A126 1.H77 1.6806 e 0 0160.0.016e 0 0172 0 0183 0 0199 0 0225 0 0307 0.1364 0 1764 0 2066 0 2326 02M3 0.2784 02995 0.3196 3500 6 77.2 1760 276.2 3791 487.6 508 4 779 4 1224 6 1338 2 1822 2 1495 5 1%3.3 1629.2 1693 6 17b7.2 s 0.1274 0.2899 0 4312 0 5585 0 6777 0.7973 0 9508 1.3242 1.4112 14709 1.5194 1.M18 1.6002 1.6353 1.669) e 0 0159 0.0164 0 0172 0.0182 0.0198 0 0223 0 0287 0 1052 0 1463 0.1752 01994 02210 0.2411 0.2601 02783 4000 6 73.5 177.2 277.1 379 8 487.7 606 5 763 0 1174.3 1311 6 1403 G I481.3 1552.2 1619A 1685.7 17504 s 01271 0.2893 0.4304 0.5573 0 6760 0 7940 0 9343 1.2754 13807 1.4461 1.4976 1.5417 1.5812 14177 14516 e 0 0159 0 0164 0 0171 0 0181 0.01 % 0 0219 0.0268 0.0591 0 1038 0 1312 0 1529 0.1718 0 18*0 0 2050 0.2203 5000 a SI 1 179 5 2791 381.2 488.1 604 6 746 0 1042 9 1252.9 13(4 6 1452.1 1529 1 16039 1670 0 1737.4 s 0.1265 0.2861 04267 0 5550 0 6726 0.7880 0 9153 1.1593 1.3207 3.4001 1.4582 1.5061 15481 1.5863 1A214 e 0 0159 0 0163 0 0170 0 0160 0 0195 0 0216 0 02 % 00397 0 0757 0.1020 0.1221 0.3391 0.1544 01684 0.1817 60C0 6 83.7 181.7 281.0 362 7 AP8 6 602 9 7361 9451 1168 8 1323 6 1422 3 1505 9 1562 0 1654 2 17242 s 0 1258 0.2670 0 4271 0 5528 06693 0 7826 0 9026 1.0176 1.2615 1.35'4 1.4229 1.4745 1.5194 1.5593 15962 e 0.0158 0.0163 0 0170 0 0180 0 0193 0.0213 0 0248 0 0334 0 0573 0 031 A 0 1004 0.1160 0.1298 0 1424 0.1542 7000 6 86.2 184 4 283 0 384 2 449 3 601 7 729 3 901.8 1824 9 1281 7 1392 2 1482.6 1%31 16396 1711.1 s 03252 02859 04256 0 5507 0 6563 0 7/77 0 8926 1 0350 12055 1.3 D I 13904 1.44u6 1.4938 1.53'5 1.5735 TABLE A.4 PROPERTIES OF SUPERHEATED STEAM AND COMPRESSED WATER (TEMPERATURE AND PRESSURE) (CONTINUED) A.6 ,.u

namesy.amers.r a, Am/ M// f ~ g[' Y/ A I J IX I steal 44/J 7A / N //17 ~ ihTb8$ I'id/Ifk t l/llbSWN / N ///%J V / 7s / / IW /. I sao ; /llrnNI I M i/l %J W / 7s // It f I anb' SfkN'N) kI / NI I 1% N/, n,f%,, ~,> / /}!,E $7 N //fYJ/ x i / g , ~. I hb'$fl W'WL 7r~i ) fMK409 %n "l (MHK7%4 M/h ~ l LM ~ MR7WYAb MKNM K//Y \\ ~ MMNM/D /Y ~ 17b7%M5WM .AMMx M ~ JMX/MM7/ ~ R...M2MV FIGURE A.5 MOLLIER ENTHALPY-ENTROPY DIAGRAM A.7 ~

PROPEL 4 TIES OF WATER I Density e (Ibsfili 8'A Temp Satursted PF) Liquid 1000 2000 2100 2200 2300 2400 2500 8000 32 82.414 62.637 62.846 62.867 62.888 62.909 42.93 82.951 63.056 50 62.38 62.55 62.75 62.774 62.798 C2.822 42.846 62.87 62A9 100 61.989 62.165 62.371 82.390 62.409 62.427 62.446 62.465 62.559 200 60.118 80.314 60.511 60.53 60.549 80.568 80.587 60.606 60.702 300 57.310 57.537 57.767 57.79 57.813 57A36 57.850 57.882 57.998 400 53.651 53.903 54.218 54.249 54.28 54.311 54.342 54.373 54.529 410 53.248 53.475 63.79 53.825 53.86 53.89 53.925 53.95 54.11 l 420 52.798 53.025 53.36 53 40 53.425 53.46 53.50 53.53 53A9 l 430 52.356 52.575 52.925 52.95 52.99 53.02 53.065 53.09 53.265 440 51.921 52.125 52.42 52.45 52.475 52.51 52.54 52.56 52.275 450 51.546 51.66 52.025 52.065 62.10 52.14 52.175 52.21 52.41 460 51.020 51.175 51.56 51.61 51.64 51.68 51.725 51.76 51A6 470 50.505 50.70 51.1 Si.14 51.175 51.22 51.25 51.30 51.50 480 50.00 50.20 50.62 50.66 50.7 50.74 50.78 50.625 51.035 4!G 49.505 49.685 50.13 50.175 50.22 50.265 50.31 50.35 50 575 500 48.943 49.097 49.618 49466 49.714 49.762 49.81 49.858 50.098 510 48.31 48.51 49.05 49.101 49.152 49.203 49.254 49.305 49.56 520 47.85 47.91 48.46 48.515 48.57 48.625 48.68 48.735 49.01 530 47.17 47.29 47.86 47.919 47.978 48.037 48.096 48.155 48.45 540 46.51 47.23 47.296 47.362 47.428 47.494 47.56 47A9 550 45.87 46.59 46.658 46.726 46.794 46 862 46.93 47.27 560 45.25 45.92 45.994 46.068 46.142 46.216 46.29 46.66 570 44.64 45.22 45.30 45.38 45.46 45.54 45.62 46.02 580 43.66 44.50 44.586 44.672 44.758 44.844 44.93 4536 550 43.10 43.73 43 825 43.92 44.015 44.11 44.205 44.68 600 42.321 42.913 43.017 43.122 43.226 43.33 43.434 43 956 610 41.49 41.96 42.08 42.196 42.314 42.432 42.55 43.14 620 40.552 40.950 41.083 41.217 41.35 41.483 41A16 42.283 41.44 630 39.53 40.388 ~ 8T491 3 640 39.26 650 37.31 38.000 660 36.01 36.52 670 34.48 34A98 680 32.744 32.144 690 30.516 TABLE A.6 PROPERTIES OF WATER, DENSITY A.8 f g

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) i 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 27 ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, WM i ANSWER 5.01 (1.00) b REFERENCE TPT TS B3.2.3 1 015/020; K5.05(3.1/3.5) ANSWER 5.02 (1.00) c REFERENCE Comprehensive Nuclear Training Operations (CNTO), pp 4-16/27 001/000; K5.13(3.7/4.0) ANSWER 5.03 (1.00) l n\\ y a REFERENCE CNTO ' Reactor Core Control' Section 4... y; s,e Moe TreuAA Gcerc. ts. kv, 4 Cert !% toei Ft9 Z-t.Tv 001/000; K5.38(3 5/4.1) i ANSWER 5.04 (1.00) b REFERENCE CNTO, ' Reactor Core Contro1*, pp 5-10 j 001/010; K5.31(2.3/3.1)

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o ATLANTA, GEOAGIA 30323 '+,.... J \\ R PLANT OPERATION, FLUIDS, AND PAGE 28 ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WM l ANSWER 5.05 (1.00) c ~ REFERENCE TPT Requal Lesson Plan, Cycle II, Day 1-1985 TPT SD13, 'CVCS', pp 23 004/020; A2.13(3.4/3.9) ANSWER 5.06 (1.00) a REFERENCE CNTO, ' Thermal / Hydraulic Principles and Applications, II', pp 10-45/48 006/050; K5.01(2.9/3.1) ANSWER 5.07 (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 roos in manual)

4) Higher (as above, delta T increases, Th increeses)

5) Same (Primary power = secondary load)

REFERENCE NUS, Vol 4, Units 1.3, 3.2 CNTO, ' Thermal / Hydraulic Principles and Applications', pp 12-15/18 002/0005 K5.01(3.1/3.4) ~ =

mn mee UNetgOSTATES 4 NUCLEAR RE2ULATORY COMMISSION o mEGloN N L' \\ 101 hlARIETT A STREET, N.W., SulTE 2000 8 ATLANTA, GEOAGIA 30323 g e ,,e THEORY OF NUCLEAlt POWER PLANT OPERATION, FLUIDS, AND PAGE 29 ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WM ANSWER 5.08 (2.00) e) Decrease (+.5 ea) b) Increase c) Increase d) Increase REFERENCE SON /WBN License Requal Training, " Core Poisons' CNTO, ' Reactor Core Control', pp 6-22/28 001/000; K5.09 (3.5/3.7) & K5.02 (2.9/3.4) & K5.10 (3.9/4.1) ANSWER 5.09 (1.50) a) Higher (+.5 ea) b) Lower c) Higher REFERENCE TPT OP 1009.1; Plant Curves 001/000; A2.07(3.6/4.2) ANSWER 5.10 (.50) decrease REFERENCE CNTO, " Thermal / Hydraulic Principles and Applications, I', pp 2-58/5' 000/027; EK1.03(2.6/2.9) ANSWER 5.11 (1.50) a) decreasei (no answer) (+.25 ea response) b) (no ans); increase c) decreasei decrease d) decrease; decrease

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{ 101ilARIETTA 8TREET.N.W SulTE 2000 ATLANTA, GEORG4A 30323 o*% f 4 C THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 30 --- vaggassvagg1Cg-------------------------------------- ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WM REFERENCE CNTO ' Thermal / Hydraulic Principles II", pp 12-39-45 039/000; A2.05(3.3/3.6) ANSWER 5.12 (1.50) o) False (+.5) b) -Post accident heating of Reference Leg (+.5 ea) -Reference Le3 leakage REFERENCE NRC IE Info Notice 84-70 (4 Sep 1984) TPT Lesson Plan for Requal Cycle II-1985 011/000; K4.03(2.6/2.9) ANSWER 5.13 (2.00) g q g) -1) Control rods within + or - 15 inches of group demand position (+.5 ea) -7) 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 B3.2.5 001/000; K5.46(2.3/3.6) ANSWER 5.14 (1.00) -Minimizes thermal. stress due to more uniform temp difference of fluids -The outlet temp ^VT the colder fluid approaches the inlet temp of the hotter fluid -A more uniform' heat transfer rate is achieved throughout the heat exchanser (+.33 eafoetwjl) l - mo f f PMic.ut REFERENCE CNTO, ' Thermal / Hydraulic Principles and Applications *, pp 5-10 004/020; K5.02(2.5/2.9)

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{g NUCLEAR REULATORY COMMISSION R.O.oN. I \\ 101 MARIETTA STREET, N.W., SUITE 2000 2 ATLANTA, GEORGIA 30323 \\...../ ~""nRY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 31 ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WH ANSWER 5.15 (2 00) a) Unit 4 (+.5) due to a lower Beta coefficient at EOL (+.5) b) Unit 3 (+.5) due.to MTC being less negative, so Tavs must decrease come to add + reactivity) (+.5) REFERENCE CNTO ' Reactor Core Contro1*, pp 3-21 & ' Fundamentals of Nuclear Reactor Physics', pp 7-31 001/0005 K5.49(2.9/3.4) & K5.10(3.9/4.1) ANSWER 5.16 (1.00) rGaove 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 5.17 (1.00) The Moderator Temperature Coefficient (MTC) becomes negative over core life tos such a degree to over-compensate for the effect of Doppler only. (+1.0) REFERENCE CNTO, ' Reactor Core Control *, pp 3-37/40 001/0003 K5.49(3.4/3.7) ~ l l

gha ne% UNITt0 STT.TES ,#p k. NUCLEAR REGULATORY COMMISSION oN. U 101 MARIETTA STREET, N.W., SUITE 2000 4 ATLANTA, GEORGIA 30323 \\,*...*/ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 32 ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, WM ANSWER 5.18 (2.00) fy (or @ gy, EAD = C(Pt - Pb)/(Pt + Pb)] x 100% (+.5) Initiallly, Greater power generated in lower segment of quadrant and EAD will be more negative. This conditior will be accentuated at xenon burns out in the lower and builds in the upper sessents. (+.75) Ac xenon builds into the lower segment while depleting in the upper section due to the neutron flux shift, a higher percentage of power will be generatch % in the upper segment and EA0 will shift towards a positive value.(+.75) REFERENCE TPT OP-12304.8; CNTO, ' Reactor Core Control', pp 4-28/29 and Section 8 001/010; K5.34(3.2/4.1) ANSWER 5.19 (1.00) Ensures design margins to core limits will be maintained (+.75) under both oteady-state and anticipated transient conditions (+.25) REFERENCE TPT TS B3.2.2 000/005; EK1.06(2.9/3.8) ANSWER 5.20 (1.00) This is due to ' Uncovering' of the sources by that bank (+.7) causing a tipproximatealy 1/2 decade increase in count rate (+.3) REFERENCE TPT OP 0202.2, Step 4.2.2 001/000; K1.05(4 X/4.4)

ska ne% UnlTED STAT 38 /p ,g NUCLEAR REZULATORY COMMISSION o o. { f 101 MARIETTA STREET, N.W., SUITE 2900 ATLANTA.EORGIA 30323 o

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 33 ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WM ANSWER 5.21 (1.00) 4, 1, 3, 2 (.25 for each switch required to put in correct order) REFERENCE TPT GET Radeon Training Lesson Plan, pp 3 068/000; K5.04(3.2/3.5) ANSWER 5.22 (1.50) e) 7-8 (+.5 ea) 2-3 (v (8J&f o(. y-{ du k HP W fler() b) c) 6-7 REFERENCE CNTO, ' Thermal / Hydraulic Principles and Applications, II', pp 7-80/91 004/020; K5.10(3.6/3.9) [g, f6'T c) p// 4(t ) ANSWER 5.23 (1.00) P pr, t unc b o for 2250 psia, sat temp = 652 (+.5) with Tavs = 575, Th = 607 (+.5) -sive +/- 2 degrees in determining Th [Qk45degreesF REFERENCE Steam Tables 001/000; K5.56(4.2/4.6) i

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[ o RE000N N 3 .I 101 MARIETTA STREET.N.W SUITE 2900 I ATLANTA,GEORotA 30323 \\, *... * / 6. PLANT SYSTEMS DE5IGN, CONTROL, AND INSTRUMENTATION PAGE 34 ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WH 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) ANSWER 6.02 (1.00) c REFERENCE TPT E0P E-1.2, 1.3, 1.4 006/020; A4.02(3.9/3.8) ANSWER 6 (1.00) / d REF ENC T SD153 ' Service and Fire Water', pp 7-9 076/000; K1.15(2.5/2.6) ANSWER 6.04 (1.00) b REFERENCE TPT SD117 'AFW', pp 10 013/000; K4.04(42374.5)

We% venTno aT:.Tas 9 NUCLEAR REIULATORY COMMISSION neosow:: a 3 I to1 MARETTA STREET.N.W sutTE 2000 I ATLANTA, GEOAGIA 30323 os.,...../ / pl. ANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 35 ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, WH ANSWER 6.05 (1 00) b REFERENCE TPT Lesson Plan on ' Gamma Metrics Neutron Detector' Requal Cycle IV-1985 ~ 015/000t K6.01(2.9/3.2) / ANSWER 6.06 (1.00) e REFERENCE TPT SD68 ' Radiation Monitoring and Protection System *, pp 11/12 034/000; K6.02(2 6/3.3) ANSWER 6.07 (1.50) o) False (+.5 ea) b) False c) True REFERENCE TPT SD63 'ESFAS', pp 32,52, FIG 14 006/000; K1.02(4.3/4.6) ANSWER 6.08 (1.00) c) Halleability-- flows into surface defects (+.5) b) False (+.5) REFERENCE TPT Requal Cycle _LV-1985, Day II TPT SD7 'RCS', Pp 37 002/000; K4.05(3.8/4.2)

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UNITED STATES NUCLEAR REGULATORY COMMISSION o neoeoN u \\ t 0188AR!TTTA ETREET. N.W., SUITE 2000 ATLANTA, GEORGIA 30333 %,...../ ^WT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 36 ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WM ANSWER 6.09 (1.25) c) Prevent reverse flow from the VCT (+.5) b) Prevent lifting 41 seal and having it hang up (+.75) REFERENCE TPT SD8 'RCPs', pp 27-28 003/000; PNG 7(3.5/3.9) ANSWER 6.10 (1.50) Q,5 ggt) (+.5)bw.vf, c) Loss of voltage sensed by 2 relays on either 4KV bus A or B b) -Manually reinitiate the SI signal upon the loss of voltage (+.5 ea)vac somd, cu -Manually restart SI equipment taking care not to overload diesel REFERENCE 7QEgrge Bus Stripping / Load Sequencing', pp 9, 16-17 P 064/000; K4.10/K4 11(3.5/4.0) .f ANSWER 6.11 ( 78.Mgg Bridge crane; C : r r d '_' 7 ; ;, 1..ii a itch: Highest obstruction in the SFP and canal (+.25 ea) REFERENCE TPT SD44 ' Fuel Handling System *, pp 16 NRC IE Info Notice 85-12, 11 Feb 1985 034/000; K4.02 (2.5/3.3) W \\ 6+TStTT {//.O ) ANSWER 6.12 -kk k Nnr-t .2 a n c ern n e a i (.d b) Closes Gas Decay Tank Discharge Valve (RCV-014) [' d c) Closes S/G Liquid Sample Isolation Valves (2800, 2001, 2002) Blowdown Flow Control Valves (6278 A, B & C) gg 3 Dump Valve to Discharge Canal (LCV-6265) u

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NUCLEAR REi'ULATORY COMMISSION g UNffSD 877.788 o namoN u \\ 101 MARIETTA STREET, N.W., SulTE 1900 ~ I ATLANTA, GEORO4A 30323 \\...../ ~ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 37 ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, WM REFERENCE TPT SD68 ' Radiation Monitoring and Protection System *, pp 34/35 000/059 & 060; EA2.05(3.6/3.9) & (3.7/4.2) ANSWER 6.13 (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 49 002/0001 K4.10(4.2/4.4) ANSWER 6.14 (1.00) 1) 862 A and B (RWST to RHR) (+.25 ea[or % 2) 864 A and B (RMST isolation) 3) 865 A, B and C (Accumulator Isolation) 4 866 A and B SI Hot Les Injection) REFE NCE h TY g Elba.CS *, pp 36 006/000; K4.08(3.6/3.7) ANSWER 6.15 (1.50) 1) Hi side of S/U Xfrar energized (+.3 ea) 2) Unit 3 C/'J X' r has been bus cleared 3) DG Breaker to 4A bus open 4) 4A to 48 bus tie open 5) 4A bus not~1ocked out FE k E TPT SD170 ' Emergency Bus Stripping / Load Sequencing', pp 6-7 W Of 4TO[ E jgE 000/055; EA2.03(3.9/4.7)

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NUCLEAR REQULATORY COMMISSION e mEneoN u 3 101 MARIETTA STREET, N.W., SulTE 2900 ATLANTA, OEoRotA 30323 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 38 ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WH ANSWER 6.16 (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 9/10 028/000; A4.03(3.1/3.3) ANSWER 6.17 (1.50) 1) RHR HXers (+.25 ea) 2) RHR Pump seal coolers 3) SI Pump oil coolers and seals 4) SFP HXers 5) CS Pump seal coolers 6) Emergency Containment Coolers REFERENCE TPT SD40

CCW',

pp 9 008/000; K1.02(3.3/3.4) l LS ANSWER 6.18 (PT00) c) Backleakage froni S/Gs via check valves (+.75) b) Vent the pump casing (+.4) once a shift (+.1) REFERENCE TPT EO #3, Cycle I Requal-1985 ] 035/010; K1.01(4 2/4.5) l l

gan se% UlstTED STAT 58 k NUCLEAR REQULATORY COMMISSION REoeon u e 5 101 MARIETTA 8TREET, N.W., SulTE 2000 ATLANTA, GEORGIA 30323

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 39 ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WM ANSWER 6.19 (1.50) o) damper trouble annunciator (+.5) b) Accordian type mechanism w/ a fusible leak (+.5) if hot gases pass thru duct, fusible link melts (at approx 165 des F) releasing the damper and sealing the duct. (+.5) REFERENCE TPT Lesson Plan ~'f or Requal Cycle IV-1985, ' Appendix R Update' 086/ , A1.04(2.7/3.3) / .uM \\ ) ( h,3) ANSWER 6.20 (2.00) l. ,p g, 0) T[tfine is runback at 200%/ min for 7.5 seconds (+ g, stop for M seconds then repeats cycle if con on still exists /.Y) b) The rod position indicating system initiates the runback (,+.5) 1 e r:; a.:

dur pu.cr as 17^% m senseu by Tvi b me 1st st;3: i*npulce prc;-

sure (+VL o f tTSGNM A A.au DKorPLO (A0060MOH cowr ACM6TF1 [f. C) REFERENCE TPT SD127 %f0* T*L* ain Turbine 'M Control *, pp 18/19 & Fi3 11 /, Jk. Li 1Kd4 045/000; K4.12(3.3/3.6) ANSWER 6.21 (1.50) o) RWST (+.25 ea) CVCS Holdup Tanks b) .iuel rack design only allows insertion in specific locations (+.5@c4t W. 2 The center-to-center distance ensures <.95 Keff for spent fuel, (+.5) (even if unbo is used) n u. E Acki CournL.r a t e d w a t eSS toss H M R T t L ANeMya REFERENCE 'S ent Fuel Pool Cooling, Purification and Ventilation', pp 4-6 TPT SD41 WT-7( r31 f71/ Para t 033/000; K4.05(3.1/3.3)

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e RE000M il { f 101 MARIETTA STREET, N.W., SulTE 2000 ATLANTA, GEORGIA 30323 o %...../ 60 PLANT SYSTEMS DESIGNe CONTROL, AND INSTRUMENTATION PAGE 40 ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, WH ANSWER 6.22 (1.75) a) Cabinet has the capacity to support up to 6 stationary gripper coils (+.5). So with 2 grou more, would overload / heat simultaneously( +. 5 ). [13(owa M c.ecsa# p s orWM/o@j the cabinet b) 125 VDC-Latching Rods 70 VDC-Holding Rods (+.5 for reasonse +.25 for correctly associating voltages) REFERENCE WRMW,-&iEEUY'cWh,M,iS~" A 001/050; PWG-1(3.6/4.1) ANSWER 6.23 (1.00) vTt)

1) you could parallel MCCs on different units (eliminate unit separation) p

( +g

2) the MCCs could not be in sync if paralleled causing circulating currents and create overload conditions

(+ 4,g REFERENCE TPT Lesson Plan 20-OL, APP B ' Replacement of 120 VAC Inverters'e pp 5 hh Yo'- Th~-lh 062/000; A3.04(2.7/2 9) ANSWER 6.24 (.75) i MSIV may not shut on a low steam flow situat.o. when required (+.5) if there is a loss of instrument air (+.25) REFERENCE TPT LER of 23 July-1985 NRC IE Notice 85-84 of 30 October 1985 035/010; K6.01(3.2/3.6)

/ %,fh, UNITED STATES / NUCLEAR REQULATORY COMMISSION I o RE080N N { f 101 MARIETTA STREET, N.W., SUITE 2000 ATLANTA, OE0A08A 30323 o %...../ 7. PROCEDURES - NORMAL, ABNORMAL, ENERGENCY AND PAGE 41 --- Essi5c65i5Ac 55aigac------------------------ ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WM ANSWER 7.01 (1.00) 0 REFERENCE TPT E0P-FR-S.1 001/010; A2.08(4.4/4.6) ANSWER 7.02 (1.00) D REFERENCE TPT ONOP 2608.1, pp 1/2 000/024; PWG-10(4.1/4.4) ANSWER 7.03 (1.00) d REFERENCE Wastinghouse background info for 1PT E0Ps, 'RCP Trip / Restart *, pp 49/50 000/074i EK3.07(4.0/4.4) ANSWER 7.04 (1.00) d g (2, O REFERENCE TPT E0P-ES-0.3/0.4 000/011; E A 2. 0 0 (-3.4 / 3. 9 )

g as%g%,

UNITSO STATSS NUCLEAR RE:ULATORY COMMISSION o neoeoN u 3 I 10t hlARIETTA STREET,N.W SUITE 2900 ATLANTA, GEORGIA 30323 \\,..... / v ""FDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 42 ^^^^RA5i5L5GIEAL E5UTR6L'~~~~~~~~~~~~~~~~~~~~~~~ ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WM ANSWER 7.05 (1.00) b REFERENCE VCS, SOP-401, Reactor Protection and Control System, p. 15 TPT SD7 'RCS", pp 27 016/000;A2.01(3.0/3.1) ANSWER 7.06 (1.00) C REFERENCE 10CFR20.5 PWG-15: Radeon Knowledge (3.4/3.9) ANSWER 7.07 (1.50) c) False (+.5 ea) b) True c) False REFERENCE W2stin3 ouse User's Guide for TPT E0Ps, pp 5-12 h ANSWER 7.08 (1.50) Q,3-for% f) 1) Tavs/ Tref Deviation Alarm ( A Er e a ) G.dd' ace 6 cuuar.f 2) Overpower Rod.Stop ouepk(. 3) RCS High DeltY T 4) OP/0T Delta T Rod Stop 5) RCS High/ Low Tavs 6) Rod Bank D Low limit Alarm REFERENCE TPT Requal Cycle II Lesson Plan-1985; NRC Generic Letter 85-05 of 1/31/85; TPT ONOP 2608.1/2; TPT SD5 ' Rod Control', pp 24 a SD7 'RCS', pp 69 004/020; PWG-10(4.3/4.5) i

,# Sa ne%q% UselTED STATBS NUCLEAR REQUtATORY COMMISSION 3 101 R$ARIETTA STREET.N.W SUITE 2900 a ATLANTA, GEORGIA 30323 \\,...../ ~ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 43 ~~~~kd65UL 55E L'E6 TR6L ~~~~~~~~~~~~~~~~~~~~~~~~ ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WH ANSWER 7.09 (1.00) % d0/ 5 Centainment pressure (+.35) > 4 psis (+.15) Centainment radiation (+.35) >"10EE5 R/Hr (+.15) t**9ern A,4t D-t O >, a9 ' r: (NT I REFERENCE Wastin,phouse backg&:ound info for TPT E0Psn " Instrumentation Accuracy'epp 11 r K3.02(3.0/3.3) & 4cr co ? 9 e wca.ho~ ' pp (6 +- te 4t* Iks 022/000; ANSWER 7.10 (2.00) 1) Ensure following personnel are dispatched with appropriate attachments: N.O. 18 N.O. 28 3rd licensed RC05 Nuclear Turbine Operator (+.8) 2) Take these items from control room and distribute! ICCS Valve keys; Electrical Penetration Room Keysi plant loss for both unitsi Plant Supervisor Logs; 2 pair of wire cutters (+1/0) 3) Proceed to TSC (+.2) ANSWER 7.11 (1.50) 1) PZR PORVs Closed (+.5 ea) 2) LTDN Isolation Valves Closed 3) Excess LTDN Isolation Valves Closed REFERENCE TPT E0P-ECA-0.0, pp 3 000/056; PWG-11(4.5/4.6) I ANSWER 7.12 (1.00) 1) Criticality.p.Lanned within 4 hours (+.25 es) 2) Baron Concentration < 300 ppm 3) Trip was from > 40% 4) Equilibrium Xenon existed prior to trip REFERENCE TPT OP-0202.2, pp 15 001/050; PWG-12(3.7/3.7)

g Wg'g UNITED STATES NUCLEAR RETULATORY COMMISSION naaeoN u

{ 101 MARIETTA 8TREET. N.W., SUITE 2000 ATLANTA, Of oAGIA 30323 o, j %4*..*f 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 44 ~~~~EkD56L6G56dL C6NTR6L'~~~~~~ ~ ~~~~~~~~~~~~~~ ~ ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WM ANSWER 7.13 (1.50) 1) Rod can't be moved (+.5 ea) 2) Rod misaligned by > 15 inches 3) Rod drop time not met REFERENCE TPT TS 3.2-3 000/001; EK3.02(3.2/4.3) ANSWER 7.14 (1.00) c 4 cI b % Sf/eu) An SI pump (+.25) takes a suction on Loop C (+.25) via the RHR system and its HXers (+.2 ) where its cooled by CCW (+.25) i

5 M ' N * # '# f'u" A L W ## N & l* 0rou p f* ' $

REFERENCE TPT ONOP-050, pp 5/6 Wr 11 6 3. sed 005/000; K3.07(3.2/3.6) ANSWER 7.15 (1.50) g 1) Sound Containment Evacuation Alarm 2) Stop CNTMT purge supply / exhaust fansh.f) isolation valves [v.7) 3) Close 4) W, i f, C"'"' Oc c c u;t i-::. c(me (vrm 415 ad4(est/w/rx4 [4. t r) 5) "-.it:r errlic;bic R e d r c,,i t e-. 'n 11 a r. d n 12 Y-REFERENCE 4Pf-OMOP-16008. 2, pp-2 /S-evoe o33.3. q q m, 034/000; PWG-11(2 0/4.1) _2

I pa seg gggyge STATES

p NUCLEAR REZULATORY COMMISSION naoosa l

ici esARIETTA STREET, N.W., SulTE 2000 ATLANTA,0EORGIA 30323 o %*...*/ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 45 ~~~~ d656EUEEEAE~E6 TR6E-~~~~~~~~~~~~~~~~~~~~~~~ R ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WH ANSWER 7.16 (1.50)for % 3 1) Unit 1/2 Cranking Diesels via Unit 3C 4KV bus (+.5 ea up to 1.5) 2) Unit 3C Bus transformer via Unit 3C 4KV bus 3) 4C 4 54 SU transformer to 3A bus n & swnd oce, nas sauv /kmJ fbJfCWi@ REFEhECE " " " T* TPT ONOP 9408.2 Y WW- //t Jo s r.wr read <,w (@/6u/ TPT SD140 'Hain Power Distribution', Fig 1 - 3C VMV At-14 ds.t TPr M nto-T o-/ff/

8) Ll-4 Au m S[d Tm @m to 4 & b 062/000) A2.05(2.9/3 9)

- tu, h ~3 A ANSWER 7.17 (1.00) 1) Heat sink (+.15 for CSF, +.1 for correct order) 2) Integrity 3) Containment 4) Inventory REFERENCE E0P-F.0, 'CSF Status Trees' ANSWER 7.18 (1.50) Core Exit T/C (+.5 ea) T-Hot RCS Subcooling REFERENCE SON ES-0.3, pp 6 TPT ES-0.2, pp-5-- EPE-074; EA1.02 (3.9/4.2)

/ #88 g UNITIO STATES h, NUCLEAR REQULATORY COMMISSION I o REGION Il I g 101 MARIETTA 8TREET.N.W., SulTE 2900 t ATLANTA, GEORGIA 30323 ~s.,*...*/ 7. PROCEDURES - NORilAL, ABNORMAL, EMERGENCY AND PAGE 46 RA656LU656AL"66NTRUL"" " "'"" "'~ ~~~~ ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WM ANSWER 7.19 (2.50) 0) -SI Annunciator ON (+.3 ea response) -SI pumps running -RHR pumps running -EDGs running b) -CNTHT Purge / Supply fans OFF -Purge Valves CLOSED -Instrument Air Bleed Valves CLOSED -Verify Control Room Ventilation Isolation c) -AFW steam supply NOVs OPEN -AFW Flow Regulator Valves OPEN REFERENCE TPT E-0, pp 4/5 000/007; PWG-11(4.4/4.5) ANSWER 7.20 (1.50) g g g) -Take Local / Remote switch to Local for Group A (+.4) -Close breaker using local pushbutton control (+.4) -Both of these controls are in the NORTH Electrical Penetration Room (+.2) -Indications of being in Local are:-Both Red and Green Breaker Lights out

REMOTE-LOCAL CNTRL SW IN LOCAL'

[pe. m A d) annunciator (+.5) .) ape g. mo4 4 // e[ REFERENCE .k 'h 4I D eb h.c hr PP 21/22, Fig 13 gr qg9 p g, g o r 10/000; A2.01(3.3/3.6) W ack yc o - % ~ ANSWER 7.21 (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 Wastinghouse background info for TPT E0Ps, 'RCP Trip / Restart' 000/009; EK3.23(4.2/4.3)

/ A*g UNITED 870.T88 h NUCLEAR REOutATORY COMMIS440N RteeoN u ? \\ 101 MARIETTA STAGET.C.C SulTE 3900 2 ATLANTA, GEORGIA 30333 \\,...../ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 47 ~ ~~~~~~~~~~~~~~~~~~~~~~~~ ~~~~ d656L6556IL C6UTR6L R ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, WM f l ANSWER

7. 2 9 M (2.00) o)

Lettering #of sub-tasks (+.5) b) Return to next step or sub-step on the left side (+.5) c) Monitor all remaining trees for RED terminus (+.5) and if not I encountered, suspend any ORP and perform the applicable FRP (+.5) REFERENCE Wastinghouse User's Guide for TPT EOPs, pp 3-11 l \\ l r l l

/ g ase % NUCLEAR.,EZULATORY COMMISSION WMD STATES E 'A AE000N N 5 101 MARIETTA STREET.N.W Sulf d 3000 8 ATLANTA, GEORGIA 30333 \\,...../ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 48 ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, WH ANSWER 8.01 (1.00) O REFERENCE TPT AP 0109.3, pp 5-6 PWG-21: Obtain/ Verify Control Procedures (3.8/4.1) ANSWER 8.02 (1.00) b REFERENCE I TPT TS 3.10 -1 103/000; PWG-5(3.1/4.1) ANSWEL 8.03 (1.00) c REFERENCE TPT TS 3.9 h 073/000; PWG-5(3.0/3.8) I ANSWER 8.04 (1.00) twN ~ d REFERENCE TPT TS 3.8i TPT IE Report 84-250-39/40 061/000; PWG-5(3.3/4.1) l

p>R#8%q%, g UNITIO STATSS of NUCLEAR REQULATORY COMMISSION e RaoeoM m 5 l 101 MARIETTA STREET.N.W., SUITE 2000 2 ATLANTA, GEORGIA 30323 %,...../ r anMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 49 ANSWERS -- TURKEY POINT 3&4 -86/02/00-DEAN, WH ANSWER 8.05 (1.00) d REFERENCE TPT TS B3.1.8 035/010; PNG-5(3.1/4.0) ANSWER 8.06 (1.50) c. FALSE (+.5 ea) f b. FALSE I c. TRUE REFERENCE Cot, TS, p. 3/4 0-1 and B 3/4 0-1 TPT TS 3.0.1 and 4.0.1 j PWG-5: TS Knowledge (2.9/3.9) l ANSWER 8.07 (1.50) c) No (+.5 ea) b) Yes c) Yes REFERENCE TPT TS 3.7.2(b); TPT LER 85-009 of May 1985 064/050; PNG-5(3.1/4.1) ANSWER 8.08 (1.50) c) 2 or 4 (+.5-aa) b) 3 c) 1 REFERENCE TPT AP 0103.4, p 5-7 PWG-14: Tagging / Clearance Procedures (3 6/4.0)

f

p# 88%gk NUCLEAR RE00LATORY COMMISSION UNITED STATSS E

o mEoeoN u 5 I 101 MARIETTA STnEET,N.W SulTE 2900 ATLANTA, GEORGIA 30323 %,...../ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PACE "3 ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, WM ANSWER 8.09 (1.00) o. 1 b. 1 c. 3 d. 3 c. 1 REFERENCE NA U1&2 TS table 6.2-1. TPT TS Table 6.2-1; TPT AP 0103.2, pp 14 PWG-23: Staffing / Activities (2.8/3.5) ANSWER 8.10 (1.00) 1) Plant Supervisor-Nuclear (PS-N) (+.5 ea) 2) Nuclear Watch Engineer (NWE) REFERENCE TPT AP 0103.4, pp 9 PNG-14: Taggin3/ Clearance procedures (3.6/4.0) ANSWER R.11 (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 PWC-23: Plant Staffing / Activities (2.8/3.5) . ~.. i i

/ g 888 gk NUCLEAR REQULATORY COMMISSION UNITED 870738 j o neoeoN u l 3 101 MARitTTA 8TREET N.W., SUITE 2000 ATLANTA, OEORGIA 30333 \\...../ f i 0 anMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 51 ANSWERS -- TURKEY POINT 384 -86/02/03-DEAN, WH ANSWER 8.12 (1.75) l f a) -Unit 3 4160 KV SWGR room (+.25 ea) l -Unit 4 l t -EDG Building Water Curtain 1 -Control Point Guardhouse l b) establish a fire watch patrol with backup equipment (+.5) within 1 hour (+.25) REFERENCE TPT TS 3.14 006/000; PWG-5(2 8/3.7) ANSWER 8.13 (1.50) l 10 mph (+.1) % ff* t**T from the so[ w or-d 6,I/,#

1) wind speed (+.4)

2) wind direction (+.3) uth or northeast (+.2)

3) precipitation is falling (+.5)

REFERENCE TPT SD50 ' Gaseous Waste Disposal System', pp 25; 071/000; PWG-7(3.2/3.7) ANSWER 8.14 (1.50) 1) Visual inspection of breaker position (+.3 ea) 2) Breaker light indication 3) Functional Test (eg. voltmeter) 4) Local (or Remote) Instrumentation 3) Annunciators REFERENCE T P T A P 010 3. 4, ' p p-~J PWG-13; Conduct / Verify Valve Lineups (3.7/4.0)

ja ne% UNITED 8TATBS k NUCLEAR RE ULATORY COMMISSION [ o REOeON il 3 I 101 MARIETTA STREET. N.W., SulTE 2900 4 ATLANTA, GEORGIA 30323 ~s../ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGF ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, WM ANSWER 8.15 (1'.50) I J J 5 E

CAF m

,m p

at:- ( + 6 ')

REFERENCE ,c,'t ut.L,;'.'? TPT ADM-313 l PWG-19: Fire Brigade Knowledge (3.4/4.2) ANSWER 8.16 (1.50) 1) Classification of the Emergency (+.5 ea) 2) Decision to notify state / local authorities } 3) Protective Action Recommendations ( l REFERENCE TPT EP 20101, pp 2 PWG-36: Facility E-Plan (2.9/4.7) ANSWER 8.17 (1.25) 1) Grab Sample (+.25 ea) 2) Effluent Monitors 3) Containment High Ra e Radiation Monitor l 4) Default Values ---(1) is the pr tred method REFERENCE TPT EP 26, pp 4/5 l P -36: Knowled e of E-Plan (2.9/4.7) S I I ANSWER 8.18 (1.50) .~_ 1) Tavs < 578.2 des F (+.4 for parameter, +.1 for setpoint) 2) P::t Pressure > 2220 psigo-3) Rx Coolant Flow > 268,500 GPM REFERENCE TPT TS pp 3.1-7 002/020i PWG-5(2.9/4.1)

7 p me%g UNITED STAiSS g /p k NUCLEAR REQULATORY COMMISSION 8 e Met 0N N f f 101 MARIETTA STMET, N.W., SUITE 2900 ATLANTA, OEOAQtA 30333

%...../

l l AnHINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 53 ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, WM 1 ANSWER 8.19 (1.50) o) 3A and 3B and either 4A or 4B 4160 KV buses (+.5 ea) b) 3 of 4 480 VAC Load Centers c) 4 of 6 Battery Chargers REFERENCE TPT TS 3.7.1 062/000;PWG-5(3.0/4.0) ANSWER 8.20 (1.00) c) In a sealed envelope in the PS-3 A fice (+.4) i b) The First Phone Number diple'd' is the PSN of fice emergency telephone (+.3) and the STA will d e called and the activation message will be i held (+.3) if t o'todialer is operational REFERENCE TPT EP ' , pp 10 P 4: Operate plant communications (3.1/3.3) ANSWER 8.21 (2.00) 1) To ensure evac ~ation of area is complete (+.5 es for any 4) u 2) To rescue injured / trapped personnel 3) To perform operations to mitigate the effect of the emergency 4) To determine nature / extent of the emergency 5) To establish cefinite personnel exclusiion boundaries REFERENCE TPT EP 20111, pp 1 PWG-36: Knowledg n f E-Plan (2.9/4.7)

jam ne% {g NUCLEAR REQULATORY COMMISSION UNIT S 870738 o

neoeoN u { 101 MAAlETTA 87AEET N.W., SulTE 2000 ATLANTA, GEOMOtA 30323 o,s,...../ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 54 ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, WM ANSWER 8.22 (1.50) o) 1 hour is the time limit (+.5) b) No operations which could cause dilution are allowed (+.5 ea) Core outlet temperature must be 10desFbelowsaturation{ REFERENCE TPT TS pp 3.4-2a 006/050; PWG-1(3.8/4.0) hrMYb ANSWER 8.23 (1.00) 1) Verify the other channels are not tripped (+.5) 2) Verify the RPI test panel turbine runback defeat switch is in the normalposition(Arm} i REFERENCE I TPT OP 12304.2, pp 3 015/020; PNG-1(3.5/3.9) l l ~..

pa me 4 UNITED ST'.TES / NUCLEAR REZULATORY COMMISSION h' o RE060 Nil 101 MARIETTA STREET, N.W., SulTE 2900 g ATLANT A, GEORGIA 30323 S.....* U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: TURKEY POINT 3a4 REACTOR TYPE: PWR-WEC3 DATE ADMINISTERED: S6/02/03 EXAMINER: DEAN, W H APPLICANT: ____________m.____________ 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 categorv anc a final grade of at least 80%. Examination papers will be picked up bin (6) hours after the examination starts. % OF CATEGORY % OF APPLICANT'i CATEGORY )ALUE CATEGORY VALUE TOTAL SCORE __ 1 1 1. PRINCIPLES OF NUCLEAR POWER PLANT OFERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW O 00 S 0^ _['_[____ _['_[_; ________ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 30.00 25.00 3. INSTRUMENT 3 AND CONTROLS 30.00

5.00 4.

PROCEDURES - NORMAL, A E: NORM A L, EMERGENCf AND RADIOLOGICAL CONTROL 120 00 100.00 TOTALS FINAL GRADE _________________% All work done on this enamination is mv own. I have neither given nor received aid. ~~~~~~~~~~~~~~ dEPL5CIUiI5~55G TURE n n J

UNITED STATES [Sa ne %. o NUCLEAR RE2ULATORY COMMISSION g O' REamM H j i 101 MARIETTA 5TREET, N.W., SutTE 2000 f ATLANT A, GEoROIA 30323 FR gg;F[ES OF NUCLEAR F0WER PLANT OPERATION, PAGE 2 1. ~~- fAERR567sd5 C57'HEdi'_YEdU5f5R'6"iLU5b~iL'5E QUESTION 1.01 (1.00) Which set of parameters below oest describes centrifugal p>mp runout conditions? a. High discharge pressure, low flow, high power demand b. High discharge pressure, low flow. low power demand c. Low discharge pressure, high flow, high power demand d. Low disenarge pressure, high flow, low power demand e. Low discharge pressure, low flow, high power demand GUESTION 1.02 (1.00) Which of the following curves (see attached page) representing Xenon concentration is correct for the given power history? GUESTION 1.03 (1.00i When performing a reactor S/U to full power that commenced five hours after a trip from full power eoviliorium conditions, a 0.5%/ min ramp was used. How would the resulting :enon transient vary if instead a 2%/ min ramp was used? a. The Menon dip for the 2*?/ min ramp woulc occur sooner and tne magnitvoe of the dip would be smaller, b. The <enon cip for the 2%/ min ramp woulo occur later and tne magnitude of the dip would be smaller. c. The :enon dio for the 2%/ min ramp would occur sooner and the magnitude of the dip would be larger. d. The aenon oio for the 2%/ min ramp would occur later and the magnitude of tne dip would be larger. QUESTION 1.04 (1.00) Which of the curves on the following page snows the expected trace on the NIS Startup recorder for eoval reactivity insertions in a suberitical reactor during a reactor startup? (xxxxs CATEGORY 01 CONTINUED ON NEXT PAGE xxxxx)

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ii"~I ( e e Y 4 f. f*. TIM ef a /. f 8 i 1,m e f e ~ l von / Nwbf f 1 ( i ................,i l 4

O UNITED ST ATES [pa af h #o, NUCLEAR REZULATORY COMMISSION 8' 'j ntosom s 3 g 101 M AmlETT A STREET. N.W., SutTE 2000 ATLANT A, otomolA 30323 F'Rhyggt S OF NUCLEAR POWER FLANT OPERATION, PAGE 3 1. --- isEEs567sAsiC5 IAEAi fEAsiFEE As6 ECui6 i[6U QUESTION 1.05 (1.50) State whether the situations below will generate the greatest tensile stress on the INNER or the OUTER wall of the reactor vessel. a) Heatup at a rate of 80 cegrees F/hr b) Incressing pressure 250 osig c) Cooldown at a rate of 50 degrees F/hr GUESTION 1.06 (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 occur, there is no operator action and roo control is in MANUAL, indicate whether the following parameters will be HIGHER, LOWER or the SAME at the end of the transtent compared to their initial valves. 1 :s 42 S/G steam pressure (0.5) 2) 63 RCS'loor. flow (0.5) 3) Tc in loop 41 (0.5) 4) Th in loop #2 (0.5) Si Nuclesr Fower (0.5) QUESTION 1.07 .l.50) For the changes listeo below (treat each one incependentiv) Inoicate whetner the moderator temperature coefficient will become MORE NEGATIVE, LE55 NEGATIVE or nave NO EFFECT. (Assume all other parameters are constant) si Neutron f l u:: Peek shifts radiall, outward to tne eoge of the core. bi Ecron concentration increases 100 ppm wntle core is at MOL. ci All rods in instead of all rods out.

xxxxx CATEGORY 01 CONTINUED ON NEXT FAGE axxxx) l 1

UNITED'STiTES 8'8Tg# 6 /

o, NUCLEAR REZULATORY COACISSION P'

REG 40N N 2 1 101 MARIETTA STREET. N.O.. SulTE 2000 ATLANT A. GEORGIA 30323 o PR6fCF[# ES OF NUCLEAR POWER PLANT OPERATION, PAGE 4 1. ~~~~iEEEE665EIit5C57 EEIT iEIE5EEE E6 ELU56'_EL6U QUESTION 1.08 (1.50.i An ECC is caleviated for a startup followins a reactor trio from 100% power equilibrium ::e n o n ( E:0L ). Indicate if the actual critical roo position will be HIGHER, LOWER or the SANE from the calculated Position for each of the following situations. Use attached curves as appropriate and treat each case indivicually. a) 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 ppm is used instead of the 1450 ppm curve. QUESTION 1.09 (.50) Does the Latent Heat of Vaporization INCREASE, DECREASE or REMAIN THE SAME ss saturation pressure / temperature of water is increased? GUE5 TION 1.10 (1.50i s '. TRUE or FAL5E: Durin3 colo plant conditions. vou would expect the COLD calibrated PZR level instrument to inoicate HIGHER than the HOT calibrated level instrument. (0.5) t' Give two olfferent conditions involvin3 the reference le3 which could result in a islse hi3h level indicstion on tne PZR level instrum.ent. (1.0) (xxxxx CATEGORY 01 CONTINUED ON NEti PAGE xxxxx) y y'

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UNITED STATES gf a etTg]o n NUCLEAR RE"UtATORY COMMISSION REGION il = e 101 MARIETT A STREET, N.W., SUITE 2000 g ATLANT A, GEORGIA 30323 [*RhyggtESOFNUCLEAR F0WER FLANT GPEFATION. FACE 5 1. --- isERR55fAARic57 REAf~fRAsiFER As5 FEUi5 FE55 =. QUESTION 1.11 (1.50) Match the heat transfer process in Column A to the equation that applies to that process in Column E:. COLUMN A COLUMN E: a. E;e t w e e n cold leg and hot leg 1. 0 = m EAT g of reactor (normal FC flow) 0=dm6T ptte y b. Actoss S/G tubes (primary to / secondary) 3. G = U AAT c. Across S/G (feedwater to steam) 4. G= m c6h 5. 0 = m il h QUESTION 1.12 (1.00.i Attached are curves for overall power defect and Doppler only power defect indicating both EOL and E:0L va lue s for a evele 1 core. These show Doppler defect becomes less negative over core life. W1-then does the overall power defect become more negative? GUESTION 1.13 (1.00) What are the two orimary factors that provide the or:ving mechanism for Natural Circulation flow? QUESTION 1.14 (1.00) List three significant heat transfer advantages of a counter flow neat exchanger over a parallel flow heat exchanger. tammaz CATECORY 01 CONTINUED ON NEXT PAGE

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I UNITED STATES g[ys ne3: o,, NUCLEAR RE^ULATORY COMMISSION REoeoN u y ,, 7 g 101 MARIETTA STREET,N.W SutTE 2900 g [ ATLANTA, GEORQ4A 30323 o Fk WCJF 5 0F NUCLEAR POWER PLANT OPERATION, PAGE 6 1. --- isEEs56isisics-sihi iE3ssFEE As5 FEUi5 FE5s GUESTION 1.15 (2.00) Unit 3 has just restarted following a refueling outage while Unit 4 is near EOL. Answer the following regarding the differences in plant response between the two units (e:: plain your answers)* a) At a steady power level of 10EE(-8) amps during a startup, equal reactivity additions are made (approximately 100 pcm). Which Unit will have tne higner steaov state startup rate? b) At 50% power, s centrol rod (100 pcm) drops. Assuming NO RUNBACK or OPERATOR ACTION, which Unit will have the lower steady state Tavg? QUE5 TION 1.16 (1.25) There are two effects that cause differential boron worth tochange over core life. List these two effects, their relative impact on differential boron worth and indicate which effect is the overriding factor. GUESTION 1.17 t1.50) ai What is the definition of 5hutdown rargin (50h)? (1.0) b) If a stuck

od e::Ists while the reactor is st power, what adjustment.

If snv, nust ce made to the SDM calculationi (0.5) GUESTION 1.19 (2.50) a. Define DNBR (0.5) b. Wnat is the limit on DNBR5 (0.5) c. 5ince the DNBR is not a directly ooservable parameter, name 5IX parameters the operator monitors and/or controls to ensure the DNBR is not violated. (1.5) (xxxxx CATEGORY 01 CONTINUED ON NEXT PAGE xxxxx)

d3% UNITED ST ATES /j o, NUCLEAR REZULATORY COMMISSION 9 F mEoioN o 9 y 101 MARIETTA STREET.N.O., SUITE 2000 ATLANTA, GEORGIA 30323 1. FR ypptES OF NUCLEAR F0WER PLANT OPERATION,' PAGE 7 --- isiss55isisili, sEif fEis5EEE is5 ECUi5 EE6s QUESTION 1 19 (1.00) Operating Procedure 0202.2, Unit 3tartupe 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 paenomenon and what is the appr o::imate change in count rate that is coserved? GUESTION 1.20 (1.00) Arrange the following types of radiation in order of penetrating power from LOW to HIGH: 1. Beta

2. Gamma

3. Neutron 4.

Alpha GUESTION 1.21 (1.00i Using the attached steam tables, what is the amount of primary subcooling at the cor e e>:it if the cressurl:er is at 2235 psig and Tavs is 575 degrees? (assume normal operating conditions) Gutsilus 1.;; '1

5r Y

UNIT 4 is just critical in the intermediste range when rod 0-4 (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 30 seconds into tnis too witharawal accident? Show vour calculations. GUESTION 1.22 (1.00) The pressiirl:er 'A' P O R 'J partially opens to a throttling position during operations at 85'4 power. Assuming a PRT pressure of 20 osia and saturation conditions in the pressurl: er correspanoing to 2240 psia, what is the qualitv of the steam on the downstream sloe of the F OR'J ) Show all calculations. t****x END OF CATEGORf 01

- a**)

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g8 88:4 UNITED ST

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NUCLEAR REluLATORY COMMIS$10N g REQ 4oN il e g 101 MARIETT A STREET,C.O. SulTE 2000 g ATLANTA. QEORQlA 30323 5 / 2. kt WJ.etSIGN INCLUDING 5AFETY AND EnERGENC1 SYSTEn3 PAGE 8 GUESTION 2.01 (1.00) Which of the following is NOT a source of water to the PRT1 a. Reactor Vessel Flange Leakoff Detector Drain b. Letdown Relief Usive RV-203 c. RCP Seal Water Return Line Relief Valve RV-392 d. Safety Injection Test Line Relief Valve RV-859 e.' Accumulator Discnarge Line Drain GUESTION 2.02 (1.00) Which of the following statements correctly describes the RHR System lineup when HOT LEG RECIRCULATION is established? (assume both trains of RHR are availabler 3. Eoth trains are useo to suppiv hot leg recirculation e::c l u s i v e l y. b. One train is used to supply hot leg rectreviation and the other train is used to continue cold les recirculation. c. Both trains supply both hot and cold les recirc simultaneously. d. One train is used for hot les rectre and the other train is put in standby (ie. recirculates from RHR Hxer outlet to RHR pump inlet). GUE5 TION 2.03 (1.00i Which of the statements belcw regaroing Unit 3 AFW pump steam supply valves on a loss of the 3A 4KV bus voltage is correct? a. Steam supplv valves frow all 3 5/Gs apen, b. Steam supply valves from A and 8 S/Gs open. c. Steam svFplv valves trom B anc C 5/Gs open. d. 5 team supply valves from A ano C S/Gs ocen. e. No steam supply valves will ocen as it takes loss of voltage on both 4KV bus 3A and 3B to cause the valves to open.

ssras CATEGORY 02 CONTINUED ON NEXT PAGE **mma) 1

UNITED ST".TES jh*"8 Cog j "o, NUCLEAR RE'l0LATORY COMMISSION 8' REQ @N il 101 MARIETT A STREET.CO. SulTE 2900 2 g o, f ATLANT A. oEORalA 30323 hfWJ.DSIGN I'NCLUDING SAFETY AND EnERGENCY SYSTEh5 PAGE 9 2. QUESTION 2.04 = (1.00) ,/ ) Which of tne following describes the NORMAL E:ACKUP source.of~ water to the lost Jo-t'he Fire Protection Fire Protection System if electrical power was System pumps? ~, a. The Backup Service Water pump takes 'a suction on tne Raw Water ~ Tanks and is 11nec up to the.f' ire Protection pump suction. b. A sPoolPiece is installed'between the Ftre Protection System and the Service Water Fump discharge. c. The Screen Wash Pump Olscharge is connected to a nearby fire hydrant using a nose. d. The. Elevated Storage Tank is lined up to provide gravity feed to the Fire Protection System by opening valve 794 (normally closed). / GUESTION 2.05 (1.00) Which of the following flowpaths describing how power is normally supplied to a typical vital instrument bus is correct? a. 480 VAC from vital bus, rectified ta 125 VDC, inverted to 120 VAC, and supplied to instrument cus. b. 480 VAC from vital bus, transformed to 120 VAC, and supplied to instrument bus. c. 125 VOC from batterv, supplied to batterv bus, inverted to 120 VAC, and supplied to instrument bus. d. 480 VAC from sital bus. rectified to 120 VDC, and si.ipplied to instrument cus. QUE5T 0N .06 (1.50) Incicate whether toe following CVCS valves will FAIL OPEN, CLOSEC or AS IS on a Loss of Instrument Air. a) Low Pressure Letdown Valve (PCV-145) bi Charging Flow Control Valve (HCV-121) c) Letoown Orifice Isolation Valves (CV 00 A, S and C) (xxxxx CATEGORY 02 CONTINUED ON NEXT PAGE ** max) f

a UNITED STATES /ga mah "o, NUCLEAR RET _ULATORY COMMISSION 8' ' 7, mEoeoN n 3 g 101 MAAlETTA STREET, N.W., SulTE 2000 [ ATLANT A, GEORGIA 30323 htbWJ.& SIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 10 2. GUESTION 2.07 - (1.00) Answer the following questions regarding the instrument air system TRUE or FALSE: a) Control valves that cross-connect the UNIT 3 and 4 Instrument Air compressors will automatically supply instrument air to the other unit if that unit's air corpressor falls, as long as the supplying unit's air pressure remains above 75 psis. b) If E:0TH 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 frcm UNITS 1 and 2. QUESTION 2.08 (2.50) Match the RCS penetrations in Column A with the appropriate RCS loop segment listed in Column E:. (column E: items may be u s e d a t. r e than once but only one response per penetration) Column A Column E: a. Excess Letdown

1) Loop A cold leg b.

P:r Surge Line

2) Leap A hot le3 c.

Alternate Cnarging

3) L:op A intermediate les o.

PZR 3 pray Line

4) Loop E: intermediate leg e.

RHF. Suction

5) Loop E: hot les si Loop C cold leg

7) Loop C hot les QUESTION 2.00 (1.00) a)

What is the ourcose of the silver coating on t.1 e Reactor 'le s s e l flange (0.5) 0-rings? b) TRUE or FALSE: If the ' REACTOR VESSEL FLANGE LEAKOFF HIGH TEMP' alarm actuates, the isolation valve on the inner 0-ring leak off line will (0.Si automatically close. (xxxxx CATEGORY 02 CONTINUED ON NEXT PAGE xxxxx)

m UNITED ST0TES gf a rsCg#) ~, NUCLEAR RE1ULATORY COMMISSION REQ 60N :: y g 2 a 101 MARIETTA STREET, N.O SulTE 2000 [ ATLANTA QEORGIA 30323 Fk WJ p[ IGN I'NCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 11 2. QUESTION 2.10 3 (1.00) Fill in the blanks in the statement below regarding the Standby 5 team Generator Feedwater Pumps (SSGFP): Besides being used during startup and shutdown, t.1e SSCFPs are also These ovmps are located adjacent to the used as a(n) __________. __________ tank and take a suction on the __________ tank. Flow to the S/Gs is controlled bv the _________ valve (s). QUESTION 2.11 (1.25) Answer the following questions regarding CVCS limitations and precautions; a) The 0::ygen concentration in tne VCT cover gas must be maintained below (.25) b) During any change in F;CS Boron concentration, at least one of what two major svstems/ components must be in operation? (1.0) GUESTION 2.12 (1.00) What is the purpose of the following precautions associated with operation of tne Reactor Coolant Fumps? a) Do not open #1 Seal Leakoff isolation valves until RCS pressure is greater than 100 psig. (0.5) o) Do not open il Seal Bypass valves untti 61 Seal Leakoff valves are open with 50 psid across il seal. (.75) GUESTION 2.13 (1.50) The sodium tetraborate decahydrate added during the injection phase after a LOCA will eventuallv be distributed by the Containment Sprav System and raise the Containment Sump pH to 8.5. What are the 2 reasons for establisling this elevated pH in the containment? (xxxxx CATEGORY 02 CONTINUED ON NEXT FAGE summa) e r

UNITED STATES a atJJg# m /

o, NUCLEAR RETULATORY COMMIS$10N 8'

7, REoion is 101 MARIETT A STREET. N.W.. SulTE 2000 j ATLANTA GEORGIA 30323 g L AWJ.O[ SIGN INCLUDING S AFETY AND ENERGENC) SYSTEMS PAGE 12 2. E QUESTION 2.14 (1.25) Describe tne complete flow path of the Containment Spray System when the RWST is too low to support spray operation, assuming spray is still required. (Identify all components, e::c ep t valves) QUESTION 2.15 (2.00) List 4 of the 5 Detign bases for the ECCS Cooling Performance following a LOCA as stated in 10CFR50.46. QUESTION 2.16 (2.00) List the o NFW Pump trips and indicate which of toe trips will cause an idle NFW pump to auto-start. (Setpoints not required) GUESTION 2.17 (1.75) a) List the 2 Emer3ency loads supplied by the service water system. (1.0) b) List 5 of the 6 normal loads supplied by the service water systen.(.75) QUESTION 2.18 (1.00) What are the 2 purposes of the interlock. that crevents the L T C'N isolation valves from opening or shutting unless all three orifice isolation valves are shut? GUESTION 2.19 (1.25) a) Freventing steam binding of AFW pumps st Turkev Point is a major concern. What is the potential cause of this steam binding? (.75) b) If AFW pump casing temperature is 150 deg F, what action is required to reduce the temperature? (Include the frequency of the action) (0.5) (xxxxx CATEGORY 02 CONTINUED ON NEXT PAGE xxxxx)

UNITED STITES [pa as fg,o, NUCLEAR REZULATORY COMMISSION 8' 'j REGION tl U a 101 MARIETT3 STREET.O.C.. SulTE 2000 f ATLANTA. GEORGIA 30323 2. Ff WJ.a SIGN INCLUDZNG SAFETY AND EMERGENCY SYSTEMS PAGE 13 QUESTION 2.20 C (1.50) a) What indication does a control room operator have that a fire damper has actuated? (0.5) b) Explain in detati the design features which allow a fire damper to auto close when required. (1.0) QUESTION 2.21 (2.00) s) Describe the runback process that occurs with the Main Turbine when the OT Delta T setpoint is e:<ceeded? (1.0> b) If. the Power Range ' ROD DROP AUTO TURBINE RUNBACK' is bypassed, what conditions must e::I s t and what system will initiate a turbine runback? (1.0) QUESTION 2.22 (1.50) a) What are the 2 sources of Barated water available for the Spent Fuel (0.5) Pool? b) What 2 design t' e s t u r e s o f the spent fuel racks ensure criticality does not occur in the Spent Fuel Pool? (1.0) <*xxxx END OF CATEGORY 02

x***)

I l'

UNITED ST'.TFS ,a ne ft 'o, NUCLEAR RE' UtATORY COMMISSION .8' RCON 11 ~ 2 I to1 MAmitTTA sTmtET.c.c, suite 2eoo [ ATLANTA, OtonGIA 30323 g NG7FyMfNTSANDCONTROLS PAGE 14 3. QUESTION 3.01 (1.00) Which of the f ollowing e:: presses the combined error s2gnal used by the Reactor Control System to generate rod motion? (Assume outward motion is represented by a '+' signal) a. (Impulse Fressure - Nuclear Power) + (Tref - Tavg) b. (Nuclear Power - Impulse Pressure) + (Tref - Tavg) (Tavg - Tref) c. (Nuclear Power - Impulse Pressure) + (Tavg - Tref) d. (Impulse Pressure - Nuclear Power) + OUESTION 3.02 (1 00) Which of the following correctly describes the actions of the AFW Flow Controllers AFTER receiving an initiation signal? a. Since the AFW flow control console Hand Indicating Controllers (HICs) ar e set to a predetermined flow rate, 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 GPh actual flow measured. c. Due to a long reset time in the flow control circuitry, the large difference between the HIC setooint and the 0 GPM actual flow measured initiallv. the valve will SLOWLi GPEN to tne 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 stabill:es, OPEN SLOWLY to the pre-set position. (Exxzz CATEGORY 03 CONTINUED ON NEXT PAGE xxxas) E' t i

ga 08 Copo, UNITED STATES NUCLEAR RE ULATORY COMMISSION J 8 "A REQlON 88 f, O 101 MARIETTA STREET O.C., SulTE 2000 ATLANTA oEOAGIA 30323 o ? = fMSJggv[NTS AN'D CONTROLS PAGE 15 3. QUESTION 3.03 (1.00) Which statement below correctly describes operation of the GAMMA-METRICS Neutron Flu:: honitor in gamma flu:. fields between 10,000 and 1,000,000 R/hr (ie. high radiation fields). a. The monitor is not designeo to operate in such high level radiation fields. b. The monitor will operate satisfactor11v in these radiattor. levels, but an adjustment should be mace to discriminate against the higher gamma flu:. c. The monitor will operate satisfactortiv, 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 f i v:: as it does in much lower radiation fields. QUESTION .04 (1.50) Indicate wnether there are 1, 2 or 3 SELECTABLE detector inputs for each of the followin9 Farameters utill:ed Dv the S/G Water Level Control System. a) S.* b Level b) Feeo Flow ci Steam pressure GUESTION 3.05 (1.00) Indicate whether the following situations would csuse the steam dump' system t: ARn ONLY, ARN i ACTUATE or HAVE NO EFFECT: a) FT-447 ilst stage impulse pressure for load reject signal) fails LOW, Mode Control in Tavs mode, Tref Tavg by 6 des F. b) Turbine Trips, Mode Control in Tavg mcde, Locc A Tavg falls HIGH, the Stesm Dump Control Reset switch is HELD in the ' Bypass' position (*x*** CATEGORY 03 CONTINUED ON NEXT FAGE

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UNITED STATES [pne%], b NUCLEAR REZULATORY COMMISSION 2 o REQlON il O g 101 MARIETT A STREET,0.0.. SulTE 2000

g f

ATLANTA, GEORGIA 30323 3. k'MSJS11dt NTS AND CONTROLS PAGE le QUESTION 3.06 (1.50) Indicate whether the following will cause a Rod Control Svstem URGENT or NON-URGENT FAILURE alarm. a) Loss of hain 100 VDC Power in tne Logic Cabinet b) Slave Cycler fails to start counting upon recolpt of 'Go' pulses c) Loose circuit caro in the Power Cabinet QUESTION 3.07 01.50) Answer the following questions regaroing ESFAS TRUE or FALSE: a) In order to generate a 'P' signal, 2 out of 3 Hi Containment Pressure OR 2 out of 3 Hi-H1 Containment Pressure signals are required. b) The S/G Delta P 'S' signal actuates wnen 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 Lcw Tavg will result in Main Feed Water control valve closure, but the main Feed Water pumps will still be running if they were cperating at the time of the reactor trip. l QUESTION 3.08 (2.00) naten the interlock descriptions in Column A w 1 *. o the appropriate logic requireo to cause rod withdrawal to be blocked in Column B. (column B ltems may be used more than once> COLUMN A COLUMN B a) Power Range High F lu:: (? 103% cower 1. 1/2 2. 2/2 l b) Overtemperature Delta T rod stop 3. 1/3 4. 2/3 c) Intermediate Range High F l u :- 5. 1/4 6. 2/4 d) Power Range Rod Drop 7. 3/4 j (***** CATEGORY 03 CONTINUED ON NEXT PAGE

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f = ~ q q

UNITED STATES ga at vq',o NUCLEAR REZULATORY COMMISSION 6 g P' mEGmNS a 2 g 101 M A AIETT A STREET, N.W., SutTE 2000 ATLANTA. GEoAGIA 30323 o RhJ$pe[NTSANDCONTROLS PAGE 17 3. = QUESTION 3.09 (1.50) Match the CVCS-related component in Column A with the appropriate immediate power svPPly in Colven B. Column A Column B a. Borte Acid Transfer Pump 3B 1. 480 VAC MCC 3A b. Charging Pump 4A 2. 480 VAC MCC 4A c. CVC3 Heat Tracing, Train A for Unit 3 3. 480 VAC MCC 3B 4. 480 VAC MCC C 5. 480 'J A C M C C 0 6. 480 VAC LC 4A ~. 430 VAC LC 3A 6. 480 VAC LC'4B GUESTION 3.10 (1.00! With the 2ressurizer level control selector switch in position III/II, cn instrument fativre causes tne following plant events in sequence (Assume no operstor actions taken>: 1. Charging flow reduces to minimum 2. Pressuricer level decreases 2. Letdown secures and heaters deenergi e 4. Level increases until high level trip Which instrument failed (II or III) snd in what direction did it fail? OUESTION 3.11 (1.00) Fill in tae blanks in the paragraph below regarcing the PHR System: PC-c00 (RHR num; A pressure detectors is interlocked with _______ and _______ to prevent opening these valves when RHR system pressure exceeds _______ psig. This interlock is to prevent overpressuri:ing the section of piping associateo with the QUESTION 3.12 (1.50) 1 s) Give the location ano the number of UV relavs that must be energi:eo j 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) l (***** CATEGORY 03 CONTINUED ON NEXT PAGE xxx*x) L

UXITED STATES [ga segIo, NUCLEAR REGULATOAY COMMISSION 3' KEDON 11 f a 101 MA lETTA STEEET, N.W., SulTE 2900 f ATLANTA. 0EOmotA 30323 fME7Fpm[NTSANDCONTROLS PAGE 18 3. QUESTION 3 13

(1.25)

Fill in the blanks below in the statement regarding Nuclear Instrumentation requirements for critical operations: The minimum number o t' operable Power Range instruments to proceed to critical operations is _____. It is reqvtred that _____ Intermediate l Range channel (s) be operable unless _____ o t' _____ Power Range full Power. channels indicate (s) QUESTION 3.14 (1.20) When nuclear power has been increased above the setpoint for permissive F-10, the operator can manually block three protective features. List the THREE features that can be blocked. QUESTION 3.15 (1.60) List the 6 reactor trips which are enabled / blocked by the reactor trip system interlock P-7. QUESTION 3 16 (2.00.' What are the 4 conditions wnich must be met for .ne overpressure mitigation system (Oh5) status lights to be ON? GUESTION 3.17 (1.00) List the 4 sets of ECCS related valves required to mitigate a LOCA whlch have their control power breakers racked out during critical operations. I I QUEETION 3.18 (2.00) si List all the Auto-Start signals for the AFW svstem. (1.2) b) Aside from opening the AFW Turbine Steam S u p p l,- valves, what other 2 actions associated with the AFW svstem occur upon an initiating signal? (0.8) (xxxxx CATEGORY 03 CONTINUED ON NEXT PAGE xxxxx) l l l

UNITED STO.TES gaat:4 d "o, NUCLEAR REEULATORY COMMISSION l' ' REQlON H h 101 MARIETTA STREET, N.W., SulTE 2000 3 ( f ATLANT A, GEORGIA 30323 3. R ET$pst NTS AND CONTROLS PAGE 19 QUESTION 3.19 - (1.50) The E:us clearin3 relavs set-up the permissive to use the startup transformer from unit 3 to supply the 4A 4KV E:vs. E:e f o r e the cross-tie breaker (3AA22) can be shut, what 5 conditions are required to be met? QUESTION 3.20 (1.75) ( a) What consequences could be expected in the Rod Control System's DC Holo l Cabinet if 2 or more 3r oups of rod drive mechanisms were placed on hold l power (excludin3 Control E: a n k D rods)? Explain your reasonin3 (1.0i l l b) Why is there both a 125 VDC and a 70 VOC power supply in the DC Hold I Cabinet? (.75) GUESTION ,3.21 (1.00) The Alternate Source Transfer Switches associated with the recently installed 120 VAC inverters have >ev locks to prevent 2 switches of the l same channel bein3 selecteo ta ALTERNATE at the same time. What are the the purposes benind this administrative key control' OUESTION 3.22 (1.00) What is the reason for usin3 a wet bulb RTD for temperature instruments wnicn provide inputs to the Reactor Pratection Svstem? l l l l (***** END OF CATEGOR( 03

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UNITED ST ATES gf@8 "%]o NUCLEAR RE2ULATORY COMMISSION g REQmN il g t01 MARiETTA STREET,C:.O SulTE 2900 2 e ATLANT A, GEORQl A 30323 hIhCgpe[ES-NORMAL, ABNORMAL, EhERGENCf AND PAGE 20 4 --- E3515E55iCEE 25siK5E------------------------ QUESTION 4.01 (1.00) The Response Not Obtaineo for the first immediate action of EOP-FR-S.1 ' Response to Nuclear Power Generation / ATW5' is to manually trip the reactor. If the reactor will not trip, then* a. Place rods in hanval and insert them into the core. b. Trio the turbine ano verify steam dumps open. c. Emergency borate the RC5. d. Dispatch operator to locally trip reactor. GUESTION 4.02 (1.00> Which of the situations below requires initiation of Emergency Boration) a. Following a R :- Trip, the rod position indicators show TWO rods which are NOT fully inserted. b. Rod Bank D Low Limit Alarm is actuated. c. An uncontrolled RCS Heatup following a reactor trip occurs. d. An u n e >:P l a i n e d decrease in reactor power occurs while at 70*. rated power. QUESTI0t1 4.03 (1.00> Which of the f ollowin 3 r e asons correctiv descrioes the basis for allowing RCP restart in E0P-FR-C.1 ' Response to Inadequate Core Cooling *. Helps to mi; tne SI flow to protect reactcr vessel from col.d water. b. Once subccoling is establishec, restsrt.n3 the RCPs helps to collapse voids that may have formed in the reactor vessel nead. c. Allows restoration of PZR pressure control ; sing normal sprays. d. Provides for cooling of the core when secondary depressurl:ation does not alleviate inadequate core cooling. (xxxxx CATEGORY 04 C0rJTINUED ON NEXT PAGE xxxxxi i

UNITED STUES p 4:ut','o NUCLEAR RE'uLATORY COMMISSION 6 RE060N 11 f, 101 MAAlETTA STREET.N.C., SulTE 2000 g ,8 ATLANT A. GEORGIA 30323 F ft Ggpud S - N'ORM AL, ABNORMAL, EMERGENCY AND PAGE 21 4. --- EA5f6C56fEAC E6sTE5C------------------------ = QUESTION 4.04 (1.00) Which of the following would cause the greatest biological damage to a man? a. 0.1 Rad of Fast Neutron. b. 1 Rem of Gamma. c. 10 Rem of Beta. d. 0.05 Rad of Alpha. QUESTION 4.05 (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 prceeeding to the newiv identified FRP. b) Unless specified, a task. need not be fully completed before proceeding to a subseocent step as long as that task is progressing satisfactorily c.' If a proceoure transition occurs, any tasks still in progress from the ~ orocevare which was in eifeet need not be cc uleted. (zazza CATEGORY 04 CONTINUED ON NEXT PAGE xxx*si t r

pa 884o9 UNITED ST.*.TES / "b, NUCLEAR RE2ULATORY COMMISSION 8' j Reason a U t 101 hlARIETTA STREET, N.C., SutTE 2000 ATLANTA, GEOAQlA 30323 o [R e ggp41[E S - N O R t'. A L, A E:N O R M A L, EMERGENCY AND PAGE 22 4. RADIOLOGICAL CONTROL r._ GUESTION 4.06 (2.00) 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 ___s___ the line voltage. The diesel generator is synchron1:ec to the grid bv observing the synchro p o i r. L e r as it moves slowly in the ___b___ direction and closing the generator breaker when the pointer is ___c___ the vertical position. The power (MW) output of the generator is then raised by adjusting the ___d___. Choose from the following: a. lower than / equal to / nigher than (0.5) b. slow / fast (0.5) c. 5 minutes to / at / 5 minutes after (0.5) voltage regulator / stator cooling (0.5) d. governor control / GUESTION 4.07 '1.00) F tll in the blanks for the fcilowing statement regarding immediate actions on a Rod Droo accident: A Reactor Trip is required if _____ or morc rods have dropped. ga Rodi e is stabill:ed, is used to control Tave._____ and when temperature 144 Initially, Tavg is controlled using tne AWM ' Tecnnical Specifications allow operation with a dropped rod for over E 5 Hours as long as oower is reduced to _____ to ensure design-margin to core limits are maintained. (xxxxx CATEGORY 04 CONTINUED ON NEXT PAGE <xxxx) l I ~

ja 4 UNITED STATES a C8: f [o, NUCLEAR RECULATORY COMMISSION y a mE0 ion il U 101 M AnlETTA STREET,0.*'J., SulTE 2000 [ Afl,ANTA, GEOn01A 30323 g FnhggpUdES-N'ORMAL, ? ABNORMAL, EMERGENCY AND PAGE 23 4. --- E3515E55iEAE EBATE5E-------------~~~~------- =- GUESTION 4.08 (1.00) Fill in the blanks in the statements below regarding startup and normal power operations: a) After leveling reactor power at _____ amps to take critical rod height data, power is increased to 2% with a steady state startup rate of no more than dpm. b) Control banks should be manipulated to adjust Tavg to within _____ deg F of Tref before shifttr.g rod control from manual to automatic. c) The reactor shall not be made critical with a oifference of greater than _____ pcm between the projected critical height and the ECC rod position. QUESTION 4.00 (1.50) List 5 possible alarms (setpoints not required) on the Main Control Board that would be indications that an inadvertant ilution were occurring while the Unit was at power. (Assume Rod Control is in MANUAL, NO Rx Trip occurs and NO oper ator actions are taken to mitigate the dilution) GUESTION 4.10 (1.50' ONOF 1003.2, ' E::c e s s i v e RCS Leakage', ioentifies numerous methods by which RCS leakage may be determined. List the 5 radiation monitor alarms which could be symptoms of RCS leakage. QUESTION 4.11 (2.00) a) After trving to identify the cause of a 5 LOW decrease in the Main Condenser vacuum, what are tne remaining three immediate actions? b) List four possible causes of a loss of Main Condenser vacuum. QUESTION 4.12 (1.00) List the two conditions (including setpoints) which determine Adverse Containment conditions. (***** CATEGORY 04 CONTIr4UED ON NEXT PAGE **x**) 9 f

,p8 '8 Ig, UNITED ST ATES 6 o NUCLEAR REULATORY COMMIS$10N 8' mEGeoM 18 2 tot MARIETTO STREET.C#a.J., SulTE 2000 [ ATLANTA, GEOmGIA 30323

g hf<GCgp41[E5-NORMAL, ABNORMAL, ENERGENCY AND PAGE 24 4.

--- gasiacaciaat CarTaat------------------------ QUESTION 4.13 (1.50) What are the Unit 3 RCO's immediate actions if the word is passed ' Fire in the control room, shift personnel report to assigned control room evacua-tion stations' ? GUESTION 4.14 (1.00i List the four valves.which reposition as a result of a Phase B ('P') isolation signal. QUE3 TION 4.15 (1.50) E0P-ECA-0.0, ' 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? GUE5 TION 4.ie (1.50) After Natural Circulation has been established, what 3 indications are monitored to cetermine RCS C00LDOWN, according to ES-0.2, ' Natural Cirev11 tion Cooldown') GUE57 ION 4.17 (3.00) List ALL the immediate action sub-steps from E-0, ' Reactor Trio or Safety I n,t e c t i o n ' t h a *. allow you to accomplish the fcllowing immediate actions

3:

Check if 5I actuated (1.2) bi Containment Ventilation Isolation (1.2) (0.6) c' Verifv AFW pumps running (zzmxx CATEGORt 04 CONTINUED ON NEXT PAGE xxxas) e

UNIIUD O s a.co NUCtEAR RE ULATORY COMMIS$10N [pa :Tg], a PCQloN il o 0 10t heARIETT A ST AEET. N.D.. SUITE 290 ATLANT2.QEoRQtA 30323 PAGE ~5 o =i a AND EhERGENCY

ABNORhAL, g

hhhGgpU[ES-NORhAL, ~~~~~- ~~~~ -- ~sA5i5C55fesC~c5 sis 5C ~~ 4. = results in level 1 50) 14%) to level control circuitrvfalse Lo-Lo level ( 4.18 signals QUESTION Grovo A PZR A failure in the Unit 4 controllers LC 459C & 46 oing false with the to operationcontrols operated and controllers. restore tnem locations of remote cont *ols are operated. the PZR heater order to anv in Heaters Indicate that any Backup in place. control room level signal the in indications any the RCP if the (1.50) to trip

cooling, is required through the core promotes 4.19 QUESTION (SBLOCA). it During a small break LOCA met.

If forceo flow are trip criteriaare the RCPs trippec. why (2 00) GUESTION regarcing E0P usage: sub-tasks which 4.20 following avestions to denote (0 51 in the proceovres Answer the is used ' Re spori s e not Obtained'suc a) What indication in seovence) be perfcrmed 'f reavireo. a (0 5 must action but CANOCT be is What oper ator required. entst7 steos in sction I? sctions ao not during performance ofa CSF Status b> contingencv contin 3enev if, ano f ur ther recuireo an ORANGE terminus on (1.C are actions c) What operstor Procedure - Recovery (Optimal a ORP is encountered? Tree symptoms cesitng witn changing plantto lowest. (1 001 4.;l for nighest priority GUESTION Arrsnge the following mechanisms from in sequence while performing an E0P, Foldout F ase Response Not ObtainedCritical Safety Function Status Tre 1* 2) 3) 0; *****) END OF CATEGORY

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END OF EXAnINATION (***** (************* f

ca 2 o V,t

1/2 at

,,t A*A' s o ,,,9 A 3 E = mc a = (Vf 1 )/t = 0.693/t1/2 2 3 an2/t1/2 XE = 1/2 av 1= II'd)) 1/2'ff

EI*1t? I
I*S )

PE = m9n v = e /t 1 I + at bl*1/2 yf =y* 2 n g,_ .gx W =, :P 1=1e 0 3 ,,y av, t.E = 931 am ,x Q = mh I=Ie = 10-x/TyL o Q = kpat I*I o 6 = UAaT TVL = 1.3/u pwr = w f HVL = -0.693/v ah p. p lo * (t) SG = 5/(1 K,ff) sur 9 = P e'l

III1
Keffx)

CA eff2} SUR = 26.06/Tp)/(IIe ff-P) CR (1 K,ff)) = CR (1

k n

x 2 lp + s j SUR = 26 SUR = 26s/t= + (s - e)T M = 1/(1 K,ff) = CA /CR, j M = (1 K,ffg)/(1 E,ffj) T = (t*/e) + ((s - eyIe] SDM = ( K,ff)/K,ff seconds T = 1/(o 8) t' = 10 I = 0.1 seconds'l T = (s - e)/(Ie) /K,ff a = (K,ff 1)/K,ff = aK,ff t (s,ff (1 + IT)] 1;d3 = Eg g d / f o = ((**/(T K,ff)) + !)d) 2, g dR/hr = ( 22 2 10) 2 (f,,g) P = (teV)/(3 x 10 R/hr = 6 CE/d g ellaneous Conversi I = eN

10ap, 1 curie = 3.7 x 10 f

Watar Parameters 1 kg = 2.21 lem 3 6tu/hr 1 np = 2.54 x 106 5tu/hr 1 gal. = 8.345 10m.1 ga]. = 3.78 liters 1 nw = 3 41 x 10 Ifn = 2 54 cm =F = 9/5=C + 32 J

7.48 gal.

3 Density = 62.410g/ft 'C = 5/9 (=F-32) 1 ft Density = 1 gn/cm3 Stu/lom 1 giu = 778 ft-lbf t >I Heat of vaporization = 970 r /10m Heat of fusion = 144 8tu14.7 psi = 29.9 in. Hg. e = 2.718 1 Atm =H O = 0.4335 lbf /in. 1 ft. 2

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m s +- -- AdB5. BU/% ' ' I m q ut w ie n y Weisme, ft'/lb 7 S-Cct:r Evep 84eem C:.ter Evep C:em Weter E,, geoem ft h "a b A A 8f 8se s G s g at 0.08859 0.01602 3305 3305 -0.02 1075.5 1075.5 0.0000 2.1873 2.1873 82 35 0.09991 0.01602 2948 2w3 3 00 1073.8 10768 0.0061 2.1706 2.1767 85 40 0.12163 0.01602 2446 146 8 03 1071.0 1079 0 0.0162 2.1432 2.1594 40 45 0.14744 0.01602 2037.7 2037.8 13 04 1068.1 1081.2 0 0262 2.1164 2.1426 45 to 0.17796 0.01602 1704.8 1704.8 18 05 1065 3 1083 4 0.0361 2.0901 2.1262 50 SO 0.2561 0 01603 1207.6 1207.6 28.06 1059.7 1067.7 0.0535 2.0391 2.0946 60 70 0.3629 0.01605 868 3 868 4 38.05 10540 1092.1 0.0745 1.9900 2 0645 70 80 0.5068 0.01607 633.3 633 3 48 04 1048 4 1006 4 0.0932 1.9426 2 0359 to 30 0.6981 0.01610 468.1 468.1 58 02 1042.7 1100.8 0 1115 1.8970 2.0086 60 300 0.9492 0.01613 350.4 350 4 68 00 1037.1 1105.1 0.1295 1A530 1.9825 100 110 1.2750 0.01617 265.4 265.4 77.98 1031.4 1109.3 0.1472 1A105 1.9577 110 130 1A927 0.01620 203.25 203.26 87.97 1025.6 1113.6 0.1646 1.7693 1.9339 120 130 2.2230 0.01625 157.32 157.33 97.96 10193 1117A 0.1817 1.7295 1.9112 130 140 2.8892 0.01629 122.98 123.00 107.95 1014.0 1122.0 0.1985 1.6910 1.8895 140 150 3.718 0.01634 97.05 97.07 117.95 1008.2 1126.1 0.2150 1.6536 1.8686 150 160 4.741 0.01640 77.27 77.29 127.96 1002.2 1130.2 0.2313 1.6174 13487 160 170 5.993 0.01645 62.04 62.06 137.97 996.2 1134.2 0.2473 1.5822 1A295 170 180 7.511 0.01651 50.21 50.22 148.00 990.2 1138.2 0.2631 1.5480 14111 180 190 9340 0.01657 40.94 40.96 158.04 984.1 1142.1 0.2787 1.5145 1.7934 100 200 11.526 0.01654 33.62 33.64 168 09 977.9 1146.0 0.2940 1.4824 1.7764 200 210 14.123 0.01671 27.80 27.82 178.15 971.6 1149.7 0.3091 1.4509 1.7600 210 212 14.696 0.01672 26.78 26.80 180.17 970J 1150.5 03121 1.4447 1.7568 212 320 17.186 0.01678 23.13 23.15 188.23 965.2 1153.4 0.3241 1.4201 1.7442 220 230 20.779 0.01685 19.364 19 381 198.33 958.7 1157.1 03388 1.3902 1.7290 230 240 24.968 0.01693 16.304 16.321 208.45 952.1 1160.6 0.3533 1.3609 1.7142 240 250 29.825 0.01701 13 802 13.819 218.59 945.4 1164.0 0.3677 1.3323 1.7000 250 ; 260 35.427 0.01709 11.745 11.762 228.76 938 6 1167.4 0.3819 1.3043 1.6862 260 l 270 41.856 0.01718 10.042 10.060 .238.95 931.7 1170.6 0.3960 1.2769 1.6729 270 290 49.200 0.01726 8.627 8.644 249.17 924 6 1173.8 0.4098 1.2501 1.6599 280 290 57.550 0.01736 7.443 7.460 259.4 917.4 1176.8 0.4236 1.2238 1.6473 290 800 67.005 0.01745 6.448 6.466 269.7 910.0 1179.7 0.4372 1.1979 1.6351 800 310 77.67 0.01755 5.609 5.626 280.0 902.5 1182.5 0.4506 1.1726 1.6232 310 320 39.64 0.01766 4.896 4.914 290.4 894.8 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 L.0990 1.5892 840 860 153.01 0.01811 2.939 2.957 332.3 862.1 1194.4 0.5161 1.0517 1.5478 860 340 195.73 0.01836 2J17 2.335 353.6 844.5 1198.0 0.5416 1.0057 1.5473 880 400 247.26 0.01864 1.8444 1.8630 375.1 825.9 1201.0 0.5667 0.9607 1.5274 400 420 305.78 0.01694 1.4808 1.4997 396.9 806.2 1203.1 0.5915 0.9165-1.5080 420 440 381.54 0.01926 1.1976 1.2169 419.0 785.4 1204.4 0.6161 0 8729 1.4890 440 460 466.9 0.0196 0.9746 0.9942 441.5 763.2 1204.8 0.6405 0.8299 1.4704 460 480 566.2 0.0200 0.7972 0.8172 464.5 739.6 1204.1 0.6648 0.7871 1.4516 480 500 680.9 0.0204 0.6545 0.6749 487.9 714.3 1202.2 0.6890 0.7443 1.4333 500 520 812.5 0.0209 0.5386 0 5596 512.0 687.0 1199.0 0.7133 0.7013 1.4146 520, 540 962.8 0.0215 0 4437 0 4651 536 8 657.5 1194.3 0.7378 0.6577 13954 540 1 SEO 1133.4 0.0221 0.3651 0.3871 562.4 625 3 1187.7 0.7625 0.6132 1.3757 540, 580 1326.2 0.0228 0.2994 0.3222 589.1 589.9 1179.0 0.7876 0.5673 1.3550 580 ' 400 1543.2 0.0236 0.2438 0.2675 617.1 550 6 1167.7 0.8134 0.5196 1.3330 100 l Sto 1786.9 0.0247 0.1962 0.2208 646.9 506.3 1153.2 0A403 0.4689 1.3092 620 ' 640 2059 9 0.0260 0.1543 0 1802 679.1 454.6 1133.7 03666 0.4134 1.2821 640 660 2365.7 0 0277 0.1166 0.1443 714.9 392.1 1107.0 02995 0.3502 1.2498 640 640 2708.6 0 0304 0.0808 01112 758 5 310.1 106e.5 0.9365 0.2720 1.2086 480 700 3094.3 0 0366 0.0386 0 0752 822.4' 172.7 995.2 0.9901 0.1490 1.1390 70C 705.5 3208 2 0.0508 0 0.0508 906.0 0 906.0 1.0612 0 1.0612 701 PROPERTIES OF SATURATED STEAM AND SATURATED TABLE A.2 WATER (TEMPERATURE) i A.3

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's y e h, s, e, e, e, e, n h e.eee6 32 018 0.01602 3302.4 3302 4 0.00 1975.5 1075 5 0 2.1872 2.1872 0 1021J e.eems e.10 35.023 0.01402 2945.5 2945.5 3 03 10738 10768 0 0061 21705 2.1766 4A3 1022A s.10 0.35 4t453 0.01602 2004.7 2004 7 13 50 1067.9 3081 4 0 0271 2.1140 2 1411 13.50 1025 7 e.15 0.20 53.160 ' S01603 15M 3 1526 3 21.22 1063 5 1984 7 00422 2 0774 2.1160 21.22 1028 3 c.20 e.30 64 48e 001604 1039 7 1039.7 32.54 1057.1 1989 7 0 0641 2 0165 2 0809 32.54 19320 e.30 0.40 72.869 0.01096 792.0 792.1 40 92 10524 1093.3 0.0799 1.9762 2 0542 40.92 1034 7 e.40 0.5 79.586 0 01607 641.5 641.5 47.62 1048 6 1096 3 0 0925 1.9446 2.0370 4742 10369 0.5 0.6 85.71S 001609 540 0 5401 53 25 1045 5 1093 7 0.1028 19186 2.0215 5324 1038 7 0.6 0.7 90 09 0 01610 466 93 466 94 58 10 1042 7 11008 0.3 18966 2.0083 58 to 1040.3 0.7 08 94 38 0.01611 411.67 411.69 62.39 1040 3 1102 6 01117 12775 1.9970 6239 1041.7 0.8 8.9 98.24 0.01612 368 41 368 43 66.24 10I8 1 1104.3 0.1264 12606 1.9870 46.24 1042.9 0.9 1.0 101.74 0 01614 333.59 333 60 69.73 10M.1 1105 8 0.1326 13455 1.9781 63.73 1044.1 1.0 3.0 126.07 0.01623 173.74 173.76 94.03 1022.1 11162 0.1750 1.7450 1.9200 94A3 1051A 2A 3.0 141 47 0.01630 118 71 118 73 109.42 1013.2 1122 6 0.2009 14854 1A864 10941 10M.7 8.0 4.0 152.96 0 01636 90 63 90 64 120 92 1006 4 1127.3 0.2199 1.6428 1A626 120.90 1060.2 4.0 6.0 162 24 0.01641 73.515 73.53 130 20 1000.9 1131.1 0.2 M 9 1.6094 13443 130.18 1063.1 5.0 6.0 170 05 0 01645 61.967 61.98 138 03 996.2 1134.2 0 2474 1.5820 13294 138Al 1065 4' 4.0 FA 176 84 0.01649 53 634 53.65 144 83 M 2.1 1136 9 0.2541 1.5587 1A168 144Al 1067.4 7.0 8.0 182 86 0 01653 47.328 47.35 150 87 988 5 1139.3 0 2676 1.5384 12060 15034 1069.2 8.0 9.0 188 27 0 01656 42.385 42 40 156.30 985.1 1841.4 0.2760 1.5234 1.7964 156.28 1070 8 9.C, 10 193.21 0.01659 38 404 3842 161.26 982.1 1143.3 0 2836 1.5043 1.7879 161.23 1072.3 30 14.696 212.00 0 01672 26 782 26 80 ISO 17 970.3 1150.5 03121 1.4447 1.7568 180 12 1077.6 84.096 15 213 03 0.01673 26.274 26.29 181.21 969.7 1150.9 0 3137 1.4415 1.7552 181.16 1077.9 15 20 227.96 0.01683 20.070 20 087 196 27 960 1 1156 3 0.3354 1.3962 1.7320 196.21 8082A 30 30 250 34 0 01701 13 7266 13 744 218 9 945.2 1164.1 03682 1.3313 1.8995 218 3 1837.9 30 40 267.25 0 01715 10 4794 10 497 236.1, 933 6 1169.8 0.3921 1.2844 1.6765 236 0 1092.1 40 50 261.02 0.01727 8 4967 8 514 250.2 923 9 1174.1 0 4112 1.2474 L6585 250.1 1095.3 80 i l 40 292 71 0.01738 7.1562 7.174 262.2 915 4 1177.6 0.4273 1.2147 1A440 262A 1088.0 00 70 302 93 0.01748 6 1875 6 205 272.7 907A 1180 6 0 4411 1.1905 1A316 272.5 1100.2 70 80 312 04 0 01757 5 4536 5 471 232.1

900.9 1183 1 0 4534 1.1675 14208 281.9 1102.1 80 90 320 28 001766 4.8777 4.895 290 7 894 6 1185.3 0 4643 1.1470 1.6113 290.4 1103.7 90 100 327.82 0 01774 4.4133 48" 298.5 848.6 1187.2 0.4743 1.1284 1.6027 298.2 1105.2 500 120 34; 27 0 01789 3 7097 112 6 877.8 1190 4 0 4919 1.0960 1.5879 312.2 1107.6 320 140 353 04 0 01803 3 2010

~ '5 0 868 0 1193 0 0.5071 1.0681 1.5752 324 5 1109.6 140 160 363 55 0 0;815 26155 2au .36 1 859 0 11951 0.5205 1.0435 1.5641 335.5 1111.2 160 180 373 08 0 01827 2.5129 2.531. J46.2 850 7 1196.9 05328 1 0215 1.5543 3454 1112.5 100 200 351 80 0 01829 2.2689 2.287 355.5 842.8 1198.3 0 5438 1.0016 1.5454 3543 1113.7 300 250 400 97 0 01665 1.8245 13432 3761 425 0 1201.1 0.5679 0 9585 1.5264 3753 '1115.8 350 300 417 3b 001889 1.5233 1.5427 394 0 806.9 1202 9 0.5682 09223 1.5105 392.9 1117.2 300 350 411 73 0 01913 1.3064 1.3255 409 8 794 2 12040 0 60M 08909 1.4968 406 6 11I8 1 350 400 444 60 00193 1.14162 1.1610 424.2 780 4 1204 6 0 6217 0 8630 1.4847 422 7 111f 7 400 450 41,6 28 0 0195 1.01224 1.0318 437.3 767.5 1204.8 06360 0.8378 1.4738 435.7 1114.9 450 500 467 01 0O!99 0 90787 09276 449 5 755.1 1204 7 06490 0 8145 1.4639 447.7 1118 3 500 550 47t 94 00199 0 82183 0 Salt 460 9 743.3 1204 3 0 6611 0 7936 1.4547 456.9 1118 6 550 400 485 20 0 0201 0 74962 0.7698 471.7 732.0 1203 7 0.6723 07738 1.4461 469.5 1116.2 600 700 .503 08 0 0205 0.63505 0 6556 491.6 710.2 1201.8 0 692R 0 7377 1.4304 488.9 1116.9 700 800 514 21 0 0209 0.54809 0 5690 509.8 689 6 11994 0 7111 0.7051 1.4163 506 7 il15.2 000 j 1 900 O ! 93 0 02i? O47965 05009 526 7 669 7 1196 4 0 7279 06753 1.4032 5232 1113.0 900 1 2000 5 '4.5 9> 0.0216 0 42435 0 4460 542.6 f 50 4 1192.9 0.7434 06476 1.3910 5306 1110.4 1000 l 1100 555 2r 0.0220 0 37af 3 0 4006 557.5 631.5 11891 0 7575 0 6216 1.3794 5531 1107.5 1100 1200 l s67.19 0 0223 034013 0.3625 571.9 613 0 1184 8 0.7714 0 5969 1.3683 5649 1104.3 1200 1300 57742 0 0227 0 30722 0.3299 585.6 594.6 1180 2 0.7843 05733 1.3577 580.1 1100 9 1300 14CD ST 7 07 0 0731 0 27871 0 3018 558 8 576 5 1175 3 0 7966 05507 1.3474 592.9 1097.1 1400 1500 5,6 20 0 0235 02b372 0 2772 611.7 550 4 1170 1 0.8035 0?283 1.3373 605 2 1093.1 1500 2000 635 60 0 02",7 0 167 % 01883 672.1 466.2 1139 3 0 8621 0 4256 1.7s81 662 6 10GS6 2000 2500 65$ 31 0 02c,6 010209 01307 731 7 361 6 1093 3 0 9139 0 3206 1.2345 714.5 1032.9 2500 3000 695 33 0 0343 0 050/3 0 0850 801 8 218.4 1070 3 0 9728 01891 1.1619 782 2 973.1 3000 3298.2 701 47 00%8 0 0 0500 906 0 0 906 0 1.0612 0 1.0612 475.9 875.9 3208J TABLE A.3 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (PRESSURE) A.4 ~~ I k

j Tesspeeches F J Abe peeen, ( gesny) 100 200 300 400 000 000 M MD 900 1000 1100 $200 lam 8400 3000 e 90161 392 5 452 3 511.9 571.5 631.1 000 7 ] 3 a 68 00 lit 0 2 189b 7 1241 S 1298 6 1936 1 1984 5 l (101.74) e 01295 20h09 2.1152 21722 2.2237 22708 2.3144 e 0.0161 18 le 90 24 102 24 114.21 126 15 13e ce 15001 161 94 173 86 ISS 79 197.70 set 42 221.53 2 6 6 48 08 8845 6 1144 8 1243.3 1288 2 1335 9 13s4 3 1433 6 1483 7 1534 7 1586 7 1639 6 1993 3 17 (162 24) s 0.1795 33716 13369 1.9943 2.0%D 2 0932 21309 2 1776 2 2159 2 2521 2.28M 2.3194 2A508 23811 2A 00161 38 84 44 93 S1 03 $7.04 43 03 69 00 74 98 80 94 86 91 92 87 95 84 les 90 110 76 116 72 30 a 68 02 1146 6 11937 1740 6 12u7.8 13355 1384 0 1433 4 14835 1534 6 1546 6 16395 IM33 1747.9 13 git;.21) s 01295 1 7926 1.8593 1.9173 1.9692 2 Clu 2 0603 2 1011 2 1394 2 1757 2 2101 2 2430 2.2744 2.30 e 0 0161 0 0166 29 899 33 963 37.985 41986 45 978 49 964 53 946 57.926 61 905 65 882 49458 73 833 77.30y 16 4 48 04 16609 !!92 5 1229 9 1287.3 1335 2 13830 1433 2 1483 4 1534 5 15405 1639 4 1403 2 17473 130 (213.03) e 0 1295 0.2940 1.8134 l.8720 1.9242 1.9717 24155 2.0563 2.0946 2 1309 2.1653 2.1982 2.2297 2.2S99 2 e 00161 0.0166 22J56 25 428 20 457 St ate 34 465 37 458 40 447 43 435 46 420 49 405 S238 55.370 30 6 68 05 164 11 1191.4 1239 2 1286 9 1334 9 1383 5 1432 9 1483 2 1S34.3 1986.3 1639 3 1083.1 174 (227.96) s 0.1295 0.2940 1.7805 1A397 13921 1.9397 1.9836 2 0244 2 0628 2.0991 2.1336 2.1665 2.1979 2.22 e 0 0161 0 0166 11 03' 12 624 I4 165 15485 17.195 18 699 20 199 21697 23 194 24489 26.183 2747 40 6 68 10 164 15 1186 6 1236 4 1285 0 13336 1382 5 1432 1 1482.5 1533.7 1505 8 1638 8 1992 7 1747 (267.25) s 0.1295 0 2940 1 6992 1.7608 13143 1A624 1.9065 1.9476 1.9860 2.0224 2.0569 2299 2.1224 2.1$16 e 0.0161 0 0156 7.257 8354 9 400 10 425 11438 12 446 13.450 14 452 15.452 16.450 17A48 18.445 1 I 40 4 68 15 165 20 1181 6 1233 5 1283 2 1332 3 1301.5 14313 14818 1533 2 1985.3 1638 4 lett e 1747.1 1 (292.71) s 0.1295 0.2939 16492 1.7134 1.7681 1A168 13412 1.9024 1.9410 1.9774 2 Alto 2.04SO 2.07H 2.1088 e 0.0l61 0 0166 0 0175 6 218 7 018 7.794 8560 9 319 10 075 10 829 11 581 12331 13Att 13 A29 14.977 30 6 48 21 16824 269 74 1230 5 12913 1330 9 1300 5 14305 1481.1 1532 6 1984.9 16380 1992.0 17463 1802 (312 04) e 0.1295 0 2939 0 4371 1 6790 1.7349 1.7542 1A289 1 8702 1.9089 1.9454 1.9000 2.0131 2.0446 2.0750 2. i e 0 0161 0.01 % 0 0175 4935 5 548 6 216 6.833 7 443 3050 8 655 9258 9 840 10Ato time 11.499 100 h 68 26 168 29 269 77 1727 4 1279 3 1329 6 41379 5 1829 7 1480 4 1532.0 1584 4 1637.4 1851.4 37et.S 19 (327.82) s 0.1295 0.2939 04371 1.6516 1.7068 1.7586 1.8036 1A451 1.8839 1.9205 1.9552 1.9003 2AIS9 2202 2.0794 e 0 0161 0 01 % 0 0175 4 0786 4 6341 S.1637 56831 6 1929 6 7006 7.2040 7.7096 S.2119 8.7130 9.2134 9 120 m GS 31 16833 265 81 1224.1 1277 4 13281 13784 1428 8 14798 15314 1983 9 1437.1 18813 1746.2 180 l (341J7) s 0.1295 0 2939 0 4371 14246 1.6872 1.7376 1.7829 1A244 1 A635 1.9001 13349 1.9040 1.9996 2 m 00 2 400; i e 0 0161 0 01 % 0 0175 3 4651 3 9526 4 4119 42585 S.2995 5 7364 6.1709 6.6036 FA349 FAS$2 FAp48 8.323:. 140 4 68 3T 168 38 265 85 1220 8 1275 3 13268 13774 1428 0 18791 15308 1983 4 1636 7 1810 9 174$ 9 (353 04) s 01295 0 2939 0 4370 1 6085 I MS6 1.71 % 1.7652 1 8071 1A461 1 8828 1.9176 1.9508 1.9825 2.0129 e 0 0161 0 0166 0 0175 3 0060 3 4413 3 6480 4 2420 4 6295 5 0132 S3945 S.7741 61522 6.$293 's.90 } 160 6 68 42 168 42 26989 12174 1273 3 13254 1376 4 1427.2 3478 4 1530.3 1582.9 1636.3 1890.5 17456 1801; (363 55) s 0 12')4 0 2938 0 4370 1.5906 14522 1.7039 1.7499 1.7919 14310 1A678 1.9027 1.9359 1.9676 1.9900 2 027; e 0 0161 0 0166 0 0174 2 6474 3 0433 3 4093 3.7421 4.1084 4 4505 4.7907 S.1299 S4457 SA014 S. 180 4 68 47 164 47 26e9/ 1213 8 1271.2 1324 0 1375 3 1426 3 1477.7 15297 1582 4 1635.9 1940 2 1745J 1801.; (373 C&1 s C 1294 01638 04370 1 5743 1 6376 1.6900 17M2 1.7784 1A176 18k45 1.0894 1.9227 1 9545 1.9849 2.014; - e 00161 0 0166 0 0174 2 3593 2.7247 3.0583 3 3743 3 6415 4 0008 4 3077 4A128 4.9165 52191 S.5 200 6 68 52 ILS $1 269 9s 12101 1269 0 1322E 1374 3 1425 5 1477.0 15291 1581.9 1435.4 1409 8 174 (351 80) s 01294 0 2935 04359 1.5593 1.6242 1677G 1.7239 1.7663 1A057 1.6426 1A776 1.9109 13427 137 e OC161 0 0165 0 0174 00166 2.1504 2 4662 2 6872 2 9410 3 1909 3 4382 34837 39278 4.1709 4.413 250 6 68 % 164 03 270 05 3/510 1263 5 13190 13716 1423 4 1475 3 1527A 1580 6 1634 4 1688 9 1744 (400 97) s 01294 0 2937 04354 0 M67 1.5951 1.6502 16976 1.7405 1.7601 18173 1A524 1A458 19177 1.9482 l e l00161 0 0165 0 3174 0 0156 1.7665 2 0044 2 2263 2 4407 2 6509 2 6585 30643 3.2688 3 4721 3.67 l 300 >' 64 79 IM 74 27u 14 375 15 12377 1315 2 1368 9 1421.3 1473 E 1526 2 15794 16333 ISS0 17434 (417.35) s 0.1294 02937 04337 C5%5 1.5703 1.6274 1.6758 1.7192 3.7591 1.7964 13317 1A652 13972 1.9 e 0 0161 0 0166 0 0174 0 018C 1.4913 I.7028 1A970 2 0332 2 2652 2 4445 2.6219 2.7980 2.9730 3.14 l 350 6 68 92 1ES S5 270 74 375 21 12515 1311 4 13M 2 1419 7 1471 8 !!?4 7 1578.2 16323 8487.1 1742 (431.73) = 01203 0 29M 0 43G7 0 5664 1.5483 14077 1A573 1.7009 1.7411 1.7787 1A141 1A477 33798 1.9i e 0 0161 0 0106 0 0174 0 0102 1 2641 1.4763 16490 18151 1 9759 21339 22901 2 4450 2 i 400 a 69 05 168 97 270 33 375 27 12451 13074 1363 4 1417 0 1470 1 1523 3 1576 9 1631.2 1 I (444.60) : 01293 02935 0 43H 0 56G3 1.5282 1.5901 1 6406 16450 1 7255 1.7632 1.7988 1A325 13647 1 e 0 0161 0 0106 0 0174 0 0186 0 9919 1.1584 13037 1.4397 15708 16972 IA256 1.9507 2A746 2. 500 h 69 32 119 19 270 51 375 38 12312 12991 1357.7 1412 7 14M 6 Ib20 3 1574 4 1629 1 lese 4 174 (457.01) s 0 1292 02934 udM4 0 5t60 14971 1 559% 16'23 1 65/8 16090 1 7371 1.7730 1.0069 13393 1 3 l! l TABLE A.4 PROPERTIES OF SUPERHEATED STEAM AND COMPRESSED WATER (TEMPERATURE AND PRESSURE) A.5 j

g,,,,,g, Tomyereten, F Clog la. (est.tssip) 100 300 300 400 600 600 700 000 000 2000 1100 1300 1380 3400 13o0 e 00161 0 0166 00174 0 0th 0 7944 0 94 % 107M i1892 13000 14093 1.5160 14711 1.7252 18284 1.93m BBB e 69 54 149 42 270 70 375 49 1215 9 1290 3 1951 8 1408 3 1463 0 1517 4 1578 9 1627.0 8482 6 1738 8 37950 ge8620) s 0.1292 0J933 04362 0 M57 14590 1.53n 15444 16351 16769 1 71 % 17517 13859 13884 13494 1379; e 00161 80866 00174 00186 00704 0 7928 0 9072 1.0102 1.1078 12023 12944 1.38 b8 1 4757 1 % 47 18,53r 700 6 48 54 589 45 270 59 375 61 487 93 1281 0 1345 6 1403 7 1459 4 1514 4 1%94 1624 8 1680 ? 1737 2 17 4: (503 08) e 01291 E2932 04MO O % 55 0 6489 35090 1.M73 16154 14580 16970 17335 1 M79 18004 I8318 1 861' e 0 0161 0 016E 0 0174 0 0186 00704 0 6774 0 7823 0 8759 O M31 1 0470 1 1249 1 2093 12825 13M9 1.444, 450 6 70 11 169 88 271 07 375 73 48734 1271 1 13392 13991 1455 m 15114 15 % 9 16227 1678 9 1733 0 1792e i (5182.). 0.1290 0 2930 0 43ts 0 %52 0 6885 14869 I5484 1.H80 16413 1 M07 13175 17522 17831 18164 1 846 e 0 0161 0 0166 00174 0 0186 0 02a4 05869 06858 07713 0 8504 0 9262 0 9998 10720 11830 12131 1.28?? 988 6 70 37 170 10 271.26 375 84 44783 1260 6 13327 1394 4 1452 2 15C4 5 1%44 1620 6 1677 1 !?34 1 1791e (531.95) s 0 1290 0.2929 0 4357 05649 06881 1.4659 1.5311 1.5422 16M3 1 M62 1.7033 1.7382 1 7783 18028 1 4321 e 0 0161 001M 00174 0 0186 0 0204 0$137 06080 0 M75 0 7603 0 8295 0 Stu 0 M22 1.ON6 1.0901 1.1521 350s 6 70 63 170 33 27344 375 M 44739 1249 3 1325.9 1389 6 1448 5 1504 4 I M I.9 1614 4 1675 3 1732 5 1790: f (544.h8) s 0.1269 0.2928 0 4355 0 % 47 O M76 1.4457 1.5149 1.H77 141M 16530 14905 172M 17589 1.7905 1.420: e 00161 0 01M 00174 0 0185 0 0203 04531 0 5440 0 6188 0 6845 0 7135 0 4121 0 5723 0 9313 0 9494 1.044i 1180 a 70 90 170.56 271 63 376 08 487.75 1237 3 1318 8 1384 7 1444 7 1502 4 1559 4 1616 3 1673.5 17310 1789: (SW.28) s 01269 02927 04353 0 5644 0.M 72 1.4259 1.49M 1.5542 1 4000 1.6410 14787 1.7141 1.7475 1.7793 1309' e 0 0161 0 01IL6 0 0174 0 0185 0 0203 0 4016 0 4905 05415 0 6250 0 6845 0 7418 C3974 0 8519 0 9055 0 956 ' 1200 4 71.16 170 7A 271.82 376 20 487 72 1224 2 1311 5 1379 7 1440 9 14494 15 % 9 16:42 1671 6 1729 4 1787J (567.19) s 0.1288 0.2926 0 4351 0.% 42 C M68 1.4061 14851 15415 1.5483 16298 I M79 1.7035 1.7371 1.7691 1.79D e 0 01(1 0 0166 00174 00185 0 0203 0 3176 0 4059 0 4712 05782 05809 06311 0 6794 0 7272 0.7737 0 819 1400 e 71.48 17:24 272 19 376 44 487 65 1194 1 12M 1 1369 3 1433 2 1493 2 1551 8 1609 9 1640 0 17M 3 1785> (587.07) s 0.1287 02923 0 4348 05636 O M59 1.3652 1.4575 1.5182 1.M70 1.60M 1.6484 1.4445 3.7145 1.7508 1.781 e 0.0161 0 0166 0.0173 0 0185 0 0202 0 0236 0.3415 0 4032 0 4555 0.5031 0 5482 0 5915 0 8388 06748 0.711 1600 6 72 21 171 69 272.57 376 69 487 60 61E 77 12794 1358 5 1425 2 1486 9 1546 6 16054 1644 3 1723.2 1782 (60487)s 01286 0 2921 04M4 0 5631 048$1 0.8129 14312 1 4963 1.5478 1.5916 1 6312 1 M78 1.7022 1.7344 1.765' e 0 0160 0 0165 0 0173 0.0185 0 0202 0.0235 0 29 % 0 3500 0 3988 0 4426 0 4436 0 5229 0.5409 0.5950 0 4?4: 1800 a 7233 172.15 272 95 376 93 487.M 615 58 3361.1 1347.2 1417.1 1440 6 1541.1 1601.2 1460 7 1720.1 1779: (4Ji/12) s 0.1284 0.2918 04341 0 5626 0 68*3 0 8109 1.4054 1.4748 1 5302 1.5753 1.41M 16528 1 8876 1.7204 1.751s t 4 e O C140 0.0165 0 0173 0.0184 0 0201 0 0233 0 2488 0 3072 0 3534 0 3942 0 4320 0 4680 0.50f 7 0.5365 0 Set. 2000 6 73 26 172 60 273 32 377.19 487 53 614 48 1240 9 1353 4 1408 7 1447.1 1536 2 1596 9 1657.0 1717.0 1777. (635 80) s 01283 02916 04337 0 % 21 0 68M 08091 13794 14578 1.5138 1.5603 14014 1.6391 1 6743 1.7075 1.738' ' e 0 0160 0.0165 0 0173 0 0184 0.0200 0 0230 0 1681 0.2293 0 7712 0.3068 0.3390 O M92 0.3000 0 4259 0 452' 2900 h 74 57 173 74 274 27 377 82 487.50 612.08 !!76 3 1303 4 13463 1457 5 1522.9 1545 9 1647A 1709 2 1770.e (668.11) s 0.1280 02910 04329 0 % 09 0 6815 0 8048 1.3076 1.4129 1.4 FM 1.5269 1.5703 1A084 1 A454 ' 16796 1.7111 e 0 0160 0 0165 0 0172 0 0183 0 0200 0.0228 0 0982 0 1755 0.2141 0.2484 0.2770 0.3033 0.3282 0.3522 0.375* 3000 6 75 83 17t S8 275 22 37847 487.52 610 08 1060 5 1267 0 13632 1440.2 1503 4 1574.8 1635 5 1701 4 17(1.4 (t95.33) s 0.1277 0.29.4 0 4320 0.5597 0 67M 0 8009 1.lM6 1.3692 1.4429 1.4975 1.5434 1.M41 1421d 1.0061 1.6448 e 0 0160 C 0165 0.0172 0.0183 0 0199 0.0227 0.0335 0 1588 0 1987 0.2301 0.2576 0.2827 OJOS% 0.3291 0.3510 3200 6 76 4 175 3 275 6 378 7 487.5 409 4 800 8 1250 9 1353 4 1433.1 15038 1570.3 1634A 1698.3 1761.2 905 C4) s 01276 0 2902 0.4317 0.5592 06768 0 7994 0 9708 1.3515 1.4300 1.48 % 1.5335 1.5/49 1 AIM 14477 1A806 e 0 0160 0 0154 00172 0 0183 0 0199 0 0225 0 0307 0 1364 01764 0 2066 0 2326 02M3 0.2784 0 2995 0.319' 3500 6 77.2 176 0 276.2 3791 487.6 608 4 779 4 1224 6 13382 1422 2 1495 5 1%33 16292 1693 6 17573 s 0.1274 02899 0 4312 0 5585 0 6777 03973 0 9508 1.3242 1 All2 1.4709 1.5194 1.H18 1.60D2 14353 1469: 0 0159 0.0164 0 0172 0 0182 0.0198 0 0223 0 0287 0 1052 0 1463 0.1712 01994 0 2210 0.2411 0.2601 0.278* 4000 6 78.5 177.2 277.1 3798 487.7 6065 763 0 1174.3 13116 1403 G 1481.3 1552.2 1619A 1485 7 1750.s e 01271 0.2e93 0.4304 0 5573 0 6760 0 7940 0.9343 1.2?S4 1.3807 1.4461 1.4976 1.5417 3.9812 14177 1.6514 e 0 0159 0 0164 0 0171 0 0181 0.01M 00219 0.0268 0.0591 0.1038 0.1312 01529 01718 018mo 02050 0220: 8300 a 81 1 179 5 2791 381.2 4 88.1 604 6 746 0 1042 9 1252.9 1364 6 14521 1529.1 16039 1670 0 1737.e s 0.1965 02861 04287 0.5550 0 6726 01880 0 9113 1.1593 1J207 1.4001 l.4582 1.5061 1.5481 1.5863 IA23d i 00159 0.0163 0 0170 C 0160 0 0195 0 0216 0 02 % 0 0367 0.0757 0.1020 0.1221 0.1391 0.1544 0.1 4000 4 83.7 181.7 281 0 362 7 4FB 6 602 9 736 1 9451 1188 8 1323 6 1422 3 1505 9 IM20 1654 2 17241 s 0 1254 0.2670 0 4271 0 5528 0 M93 0 7826 0 9026 1.0176 1.2615 1.3M4 1.4229 1.4743 1.5194 1.5593 1596J e 0.01to 0.0163 0 0170 0 0180 0 0193 0 0713 0 0248 0.0334 0 0573 0 CSI A 0.1004 01160 0.l298 0 1424 0.1542 7000 4 86 2 184 4 283 0 3464.2 489 3 401 7 729 3 901.8 life 9 1281 7 1392 2 1482 6 1563 1 1638 6 1711.1 e r)1252 0 2859 04256 05507 06043 0 7/77 0 8926 10350 12055 1.31)I 1 3904 1.4496 149m 1.53'5 1.573! TABLE A.4 PROPERTIES OF SUPERHEATED STEAM AND COMPRESSED j WATER (TEMPERATURE AND PRESSURE) (CONTINUED) i A.6 1 I ~

,, p.ti ra, &.,....J.',/ hJ by)f' 88 18 as 34 M' hh5,$, ~~ ~ ,fh7% , n, ~,, 4,. 2 i ~ i yyyyy\\/ N ///% '

# M f

/ ad' / N ///%j .aac sano kI I N/ / y '~' ) i N///%) nsso l ' NI / im (virli l / g l b R~j '7 _ saan ?gy $) h Af27t o .non &f7y 1 r ggy 9,Mypy s.so / Yg, y,{yy ' / n/ g /fD&,/ Kf4W)$ NE W Hgoy // UbVXWm/// MXMgxyn/ l f@%WW8 / \\ JN'wr.ecoie,r i FIGURE A.5 MOLLIER ENTHALPY-ENTROPY DIAGRAM \\ A.7 ~ -.. E L ....[.

I o i PROPENTIES OF WATER Density e pallt') PIIA Temp Saturated (*F) Liqui 4 1000 2000 2100 2200 2300 2400 2500 8000 32 62.414 62.637 42.846 62.867 62.888 82.909 42.93 82.961 63.056 50 62.38 62.55 82.75 42.774 62.798 62.822 82A46 62.87 42A9 100 61.989 82.185 62.371 62.390 82.409 S2.427 62.446 42.485 42.559 200 60.118 80.314 60.511 40.53 80.549 40.568 80.587 60.606 80.702 300 57.310 57.537 57.767 57.79 67.413 57.436 67.859 57282 67.998 400 63.651 63.903 54 218 64.249 54.28 54.311 64.342 64.373 M.529 410 53.248 53.475 63.79 S3.825 63.86 53A9 63.925 63.95 64.11 420 52.798 53.025 53.36 63.40 53.425 63.46 53.50 63.53 63AS 430 52.356 52.575 52.925 52.95 52.99 53.02 63.065 E3.09 63.265 440 51.921 52.125 62.42 52.45 62.475 62.51 62.54 62.58 62.275 450 51.546 51.66 52.025 62.065 62.10 62.14 62.175 52.21 42.41 460 51.020 51.175 51.56 51.61 61.64 61.48 St.725 61.76 51A6 470 50.505 50.70 51.1 Si.14 51.175 61.22 51.25 61.30 51.50 480 50.00 50.20 50.62 60.66 50.7 80.74 60.78 60.825 51435 4DO 49.505 49.685 50 13 60.175 60.22 50.265 80.31 30.35 80 575 500 48.943 49.097 49.618 49 866 49.714 49.782 49.81 49.858 s0.088 510 48.31 48.51 49.05 49.101 49.152 49.203 49.254 48.306 49.54 520 47.85 47.91 48.46 48.515 48.57 48.625 48.88 48.735 4341 f 530 47.17 47.29 47.86 47.919 47.978 48 037 48.096 48.155 48.45 540 46.51 47.23 47.296 47.362 47.428 47.494 47.54 47A9 550 45.87 46.59 46.658 46.726 46.794 46 382 44.93 4737 560 45.25 45.92 45.994 46.068 46.142 48.216 46.29 44A6 l 570 44.64 45.22 45.30 45.38 45.46 46.54 45.82' - 46.02 580 43 86 44.50 44.586 44.672 44.758 44.844 44.93 45.36 590 43.10 43.73 43.825 43.92 44.015 44.11 44.205 44.68 600 42.321 42.913 43.017 43.122 43.226 43.33 43.434 43256 810 41.49 41.96 42.08 42.196 42.314 42.432 42.56 43.14 820 40.552 40.950 41.083 41.217 41.35 41.483 41A16 42J83 41.44 630 39.53 40.388 640 38 491 30.26 650 37.31 38.006 660 38.01 Sta2 870 34.48 34A38 680 32.744 32.144 690 30.516 TABLE A.6 PROPERTIES OF WATER, DENSITY A.8

9 (" UNITED 870.TES /gx818:09'o, NUCLEAR REGULATORY COMMISSION ( y* AEoeoN H 3 y 101 MARIETTA 8TAEET. N.W.. SutTE 2900 ATLANTA, GEORGIA 30323 P'RIlifJP[ES OF NUCLEAR POWER PLANT OPERATION, PAGE 26 1. --- isEEs55isAsiCi 'sEAi iEAsiFEE As5 FE0i5 FE5p ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WH ANSWER 1.01 (1.00) c REFERENCE NUS, Vol 4. pp G-8 Nuclear Pwr Plant Operator Trns Prem, HTFF and Thermo, Sect 2E CNTO, ' Thermal / Hydraulic Princples and Applications, II', pp 10-43/44 App A: Pumps / Centrifugal, Operating at Runout Conditions (3.1/3.1) ANSWER 1.02 (1.00) c REFERENCE Comprehensive Nuclear Training Operations (CNTO), pp 4-16/27 001/000; K5.13(3.7/4.0) ANSWER 1.03 (1.00) Y kW REFERENCE CNTO ' Reactor Core Control' Section 4... U :.nkhe Twr. 4 6.e,.k tu n**q a ~ 4 Ce4 9 % vo r t 9 I-C D' 001/000; K5.38(3.5/4.1) i ANSWER 1.04 (1.00) i d REFERENCE CNTO, ' Fundamentals of Nuclear Reactor Physics', pp 8-54/55 t l

e UNITED STf,TES pm "Qq <= ,f k NUCLEAR RE ULATORY COMMISSION REoioN il e o 2 I 101 MARIETTA 8TREET, N.W., SUITE 2900 a ATLANTA OEoRGIA 30323 T 1. FRIpppt S OF NUCLEAR POWER PLANT OPERATION, PAGE 27 --- isiss55isAsiCi-.siEi iEEsifis An5 FEUi6 FE5E ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WH 015/000; K5.06(3.4/3.7) ANSWER 1.05 (1.50) c) Outer b) Inner c) Inner REFERENCE NUS, Vol 4, Unit 10.1 CNTO, ' Thermal / Hydraulic Principles and Applicationt', pp 13-57/58 004/000; K5.09(3.7/4.2) ANSWER 1.06 (2.50)

1) Lower (Higher Stm Flow >> P stm 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 12-15/18 002/000; K5.01(3.1/3.4) ANSWER 1.07 (1.50) o) More Negative (+0.5 ea) b) Less Negative c) More Negative REFERENCE Westinghouse Nuclear Training Operations, pp. I-5.6 - 16 CNTO, ' Reactor Core Control', pp 3-16/28 001/000; K5.26(3.3/3.6)

1 0 a n%q UNITED STATES [(ja k NUCLEAR REGULATORY COMMISSION o REG 40N H 2 3 101 heARIETTA STREET, N.W., SulTE 2000 4 ATLANTA, GEORGIA 30323 o, 1. RIWpIPt S OF NUCLEAR POWER PLANT OPERATION, PAGE 28 TU5RU66YUd 5C5~~5i55T TRI 5F5R d 6~ FLU 56~EL6E ~ ~ ~~~~ ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WH ANSWER 1.08 (1.50) c) Higher (+.5 ea) b) Lower c) Higher REFERENCE TPT OP 1009.1; Plant Curves 001/000; A2.07(3.6/4.2) ANSWER 1.09 (.50) decrease REFERENCE CNTO, ' Thermal / Hydraulic Principles and Applications, I*, pp 2-58/59 000/027; EK1.03(2.6/2.9) ANSWER 1.10 (1.50) a) False (+.5) b) -Post accident heating of Reference Leg (+.5 ea) -Reference Leg leakage REFERENCE NRC IE Info Notice 84-70 (4 Sep 1984) TPT Lesson Plan for Requal Cycle II-1985 011/000; K4.03(2.6/2.9) ANSWER 1.11 (1.50) c. 1 (or 5) (0.5) b. 3 (0.5) c. 5 (0.5) REFERENCE General Physics, HT & FF, pp. 180 and 181 CNTO, ' Thermal / Hydraulic Principles and Applications, I', pp 4-61/64 & 5-47

8 i >* 8%q UNITED STATES ,.f k NUCLEAR RETULATORY COMMISSION e o atosoN il 3 a 101 MARIETTA STREET. N.W., SUITE 2000 a ATLANT A, GEoAGIA 30323 o, hgippfSOFNUCLEARPOWERPLANTOPERATION, PAGE 29 __________________________________'FEUE6"EL6U TUER 66YU5UEC5~~_UEdT TRdU5EER 5Ud ~ ~ ~~~~ ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WM 001/000-K5.45 (2.4/2.9) ANSWER 1.12 (1.00) The Moderator Temperature Coefficient (MTC) becomes negative over core life tos such a degree to over-compensate for the effect of Doppler only. (+1.0) REFERENCE CNTO, ' Reactor Core Control', pp 3-37/40 001/000; K5.49(3.4/3.7) ANSWER 1.13 (1.00) 1) Density difference between cold and hot leg (+.5 ea) of k Ah *"" * '* k 2) Height difference between hot and cold legs (or S/G and Core) REFERENCE CNTO, ' Thermal /Hvdr II', pp 14-16/17 s1, m.ros nrws (b et bm*a u,1,i c P r i n c i p l e s a n d A p p l i c a t i o n s

h, ca 2, p9afn 002/000; K5.10(3.5/3.9)

ANSWER 1.14 (1.00) -Minimizes thermal stress due to more uniform temp difference of fluids -The outlet temp of the colder fluid approaches the inlet temp of the hotter fluid -A more uniform heat transfer rate is achieved throughout the heat exchanser ( +. 33 e a ) Ac aw 1) - mere e moea+ REFERENCE CNTO, ' Th e r it a l / H y d r a u l i c Principles and Applications', pp 5-10,3f/40 004/020; K5.02(2.5/2.9)

UNITED STATES / p*T&q[g NUCLEAR RESULATORY COMMISSION .? e nEasow n I 101 MARIETTA STREET;N.W., SUITE 2000 I ATLANTA, oEORGIA 30323

TypJP[ESOFNUCLEARPOWERPLANTOPERATION, PAGE 30

--- isEss55isAsics? sEAi isAssFEs As5 FEui5 FC5s ANSWERS -- TURKEY POINT 344 -86/02/03-DEAN, WH ANSWER 1.15 (2.00) a) Unit 4 (+.5) due to a lower Beta coefficient at EOL (+.5) b) Unit 3 (+.5) due to HTC being less negative, so Tavs must decrease come to add + reactivity) (+.5) REFERENCE CNTO ' Reactor Core Control', pp 3-21 & " Fundamentals of Nuclear Reactor Physics', pp 7-31 001/000; K5.49(2.9/3.4) & K5.10(3.9/4.1) ANSWER 1.16 (1.25)

1) As the fuel burns out, less boron is required, which increases the boron worth (+.5); (or less boron means decreased flux hardening and a higher effective baron absorption cross-section so worth incre ses)

2) Fission products build UF, decreasing the boron worth (+.5)

[This is the overriding effect (+.25{ N -- REFERENCE SON /WBN License Cert Trns, ' Core Poisons', pp 4 CNTO, " Reactor Core Control', p p 5-15 /16, (s). e/> A, c vee r r A - 7/27 001/000; K5.09(3.5/3.7) ANSWER 1.17 (1.50) a) 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 SDM calculation) (+.5) 4 REFFRENCE deme u s e mip e n d. TPT T/S, pp 3.2 2/3 TPT OP 1009.3 000/005i EK1.05(3.3/4.1)

>*88To UNITED STCTES gf NUCLEAR REQULATORY COMMISSION REG 60N il 2 y, g 2 3 101 MARIETTA STREET, N.W., SulTE 2000 ATLANTA, GEORQlA 30323 o 1. F%IWp;FtES OF NUCLEAR POWER PLANT OPERATION, PAGE 31 --- isEER55isisiE5 7sEAi isissFEE As5 FEUi5 FE6s TURKEY POINT 3&4 -86/02/03-DEAN, WH ANSWERS ANSWER 1 18 (2.50) c. DNBR = Heat flux (power) to cause DNB / actual heat flux (power) (0.5) b. Greater than or equal (0.2 pts) to 1.3 (0.3 pts) (0.5) c. (any 6 at 0.25 pts each) (1.5) 1. RCS pressure 2. RCS temperature 3. RCS flow 4. Rx power 5. AFD 6. QPTR 7. Rods (sequencing, overlap, position) CONSIDER OTHERS ON CASE-BY-CASE BASIS REFERENCE VCS, TS, pp. 3/4 2-15 and B 3/4 2 5 and B 2-1 and General Physics, HT & FF, p. 243 CNTO, " Thermal / Hydraulic Principles and Applications, II", pp 13-20/24 015/020; K5.09(3.5/3.7) ANSWER 1.19 (1.00) This is due to ' Uncovering' of the sources by that bank (+.7) causing a cpproximatealy 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) ANSWER 1.20 (1.00) 4, 1, 3, 2 (.25 for each switch required to put in correct order) REFERENCE TPT GET Radeon Training Lesson Plan, pp 3 068/000; K5.04(3.2/3.5)

jka me: UNITED STATES f k NUCLEAR RECULATORY COMMISSION y e RE000N 11 5 g 10t hlARIETTA STREET,N.W..SutTE 2900 ATLANTA, GEORGIA 30323 t hAIllfJPESOFNUCLEARPOWERPLANTOPERATION, PAGE 32 1. --- isEER567AARfC5, sEsi isissFEs Es5 FEUi5 FE5s ANSWERS -- TURKEY FOINT 3&4 -86/02/03-DEAN, WH g [oD N ut) ANSWER 1.21 (1.00) fer 2250 psia, sat temp = 652 (+.5)4 Tb with Tavs = 575, Th = 607 (+.5) sive +/- 2 degrees in determining Th [c/9) 45 degrees F REFERENCE h'Uw,kb W 001/000; K5 56(4.2/4.4) ANSWER 1.22 (1.25) At the end of 60 seconds: 32 SPM x 5 pmm/in x 5/0 in/ step x 1 - 100 pcm (+.25) The reactivity addition rate: 5 pcm/in x 32 SPM 518 'i n/ s t e p = 100 pcm/ min h4'p)/(Eeff-p) (+.5) The equation to be used is SUR = 2 The answer is.55 DPH (+ REFERENCE CNTO, 'Fu entals of N> clear Reacstor Physics', pp 7-71/72 01; EK1.02(3.6/3 9) ANSWER 1.23 (1.00) at 2240 psia, h3= 1115 BTU /lb (+.5 for h determination) et 20 psia, for saturation conditions, hs = 1156 BTU /lb & hf = 196 from a Hollier Diagram, moisture content is approximately 5% (95% quality) .043 >> 95.7% quality calculating:(1156-1115)/(1156-196) = (+.5 for quality determination) REFERENCE Steam Tables and Hollier Diagram 010/000; K5.02(2.6/3.0)

pa as T4,9 UNITED STATES / NUCLEAR RE ULATORY COMMISSION 2 o REG 40N 11 3 I 101 MARitTTA STREET, N.W., SulTE 2000 'g ATLANTA, GEORGIA 30323 I F2.4WJ.S[ SIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 33 2. ANSWERS -- TURKEY P_0 INT 314 -86/02/03-DEAN, WM ANSWER 2 01 (1.00) O Or 0-REFERENCE TPT SD9 'PZR and Pressure Relief', pp 37 M rA d, genac Cb (h( [h M49p g ui N a ge, Disposal *, Fi3 2 007/000; A3.01(2.7/2.9) ANSWER 2.02 (1.00) c REFERENCE TPT E0P E-1.2, 1.3, 1.4 006/020; A4.02(3.9/3.8) ANSWER 2.03 (1.00) Xb REFERENCE TPT 'AFW', fig 7 c (a ck c6 #d+" d M I O r#7c.SD117 cu c.n w cr+<r 061/000; K4.09(3.7/4.1) ApSWER 2.04 (1.00) (),. Ll>

REFERENCE' TPT SD153 ' Service and Fire Water', pp 7-9 076/000; K1.15(2 5/2.6) 1 r

i gan maev UNITED STATES g*p [g NUCLEAR RE';ULATORY COMMIS810N e nEOsoN N f 101 MARIETTA STREET, N.W., SUITE 2000 t ATLANTA, GEORQlA 30323 2. f.4W7.Gt IGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 34 ANSWERS -- TURKEY P0 INT 3&4 -86/02/03-DEAN, WH ANSWER 2.05 (1.00) o REFERENCE VEGP, Training Text, Volume 8, p. 16b-3 and Fig. 16b-1 &6 VCS, GS-2, Safeguards Power System, pp. 27 & 28 TPT DWG 5610-T-E-1592 062/000; K4.09(2 4/2.9) ANSWER 2.06 (1.50) o) Open (+.5 ea) b) Open c) Closed REFERENCE NA NCRODP 88.3, 'CVCS' SONP AOI-10A, pp 3-5 TPT SD13 'CVCS". pp 18, 21, 37 000/065; EA2.00(2.9/3.3) ANSWER 2.07 (1 00) c) True (+.5 ea) b) False REFERENCE TPT SD155 ' Plant Air Systems', pp 6/7 078/000; K4.02(3.2/3.5)

/*yn me-oq\\ UNITED STATES NUCLEAR RE ULATORY COMMISSION [ RE040N il 5 g 101 MAR 6ETTA STREET,N.W..SulTE 3900 ATLANTA, GEORGIA 30323 ellJ.G[ SIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 35 2. CNSWERS -- TURKEY POINT 384 -86/02/03-DEAN, WH ANSWER 2.08 (2 50) c) 3 b) 5 c) K% g(vua d.e 1(V* O ) 6 d) ay REFERENCE Ferley SD, 'RCS', Fig 7 NA NCRODP, 'RCS'i "ESF-ECCS'i 'CVCS'; 'RHR' TPT SD7 'RCS', pp 65-67 ; M n/J-r-i-WJs 002/000; K1.06(3.7/4.0); K1 09(4.1/4.1); K1.08(4.5/4.6) ANSWER 2 09 (1 00) c) Halleability-- flows into surface defects (+.5) b) False (+.5) REFERENCE TPT Requal Cycle IV-1985, Day II TPT SD7 'RCS*, pp 37 002/000; K4.05(3 8/4.2) ANSWER 2 10 (1 00) Bcckup to AFWi Unit 4 Condensatei Demin Water Storage; FWRV Bypass (+.25ea) REFERENCE TPT Requal Cycle IV-1985, Day 3 TPT SD112 ' Condensate and Main Feedwater', PP 6/7 l 059/000; K1.03(3.1/3.3) l ANSWER 2 11 (1 25) e) 5 (+.25) l b) RHR systemi RCPs (+.5 ea) l

I jh8884 UNITED STATES 9 NUCLEAR REGULATORY COMMISSION .t-o nE0 eon N 3 \\ 101 MARIETTA 8TREET N.W., SUITE 2000 I ATLANTA, GEOR08A 30333 o

eyJ.4)[ SIGN INCLUDING S AFETY AND EMERGENCY SYSTEMS PAGE 36 ANSWERS -- TURKEY P_UINT 3&4

-86/02/03-DEAN, WH REFERENCE TPT SD13 'CVCS*r pp 46-51 004/020; PNG 7(3.4/4.1) ANSWER 2.12 (1.00) c) Prevent reverse flow from the VCT (+.5) b) Prevent lifting ti seal and having it hang up (+.75) REFERENCE TPT SD8 'RCPs'r pp 27-28 003/000; PWG 7(3.5/3.9) ANSWER 2.13 (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 11 026/000; K4.02(3.1/3.6)

ANSWER 2.14 (1.25) Sump >> RHR Pump >> RHR HX)> CS Pump >> Spray Rings >> Sump REFERENCE TPT SD25 ' Containment Spray', pp 13e Fig 1 006/020; K4.03(3.4/3.6)

pn neev UNITED STATES ,f k NUCLEAR RECULATORY COMMIS810N 3 o REGON N 3 g 101 MARIETTA STREET. N.W., SUITE 2000 I ATLANTA, GEORGIA 30323 g .','!,0fSIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 37 ANSWERS -- TURKEY P_0 INT 3&4 -86/02/03-DEAN, WM ANSWER 2.15 (2.00) 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) P ovides for long term decay heat removal REFERENCE 10CFR50.46 TPT SD21 *ECCS', pp 5 006/050; PWG 4(4.2/4.3) ANSWER 2.16 (2.00) o, w 6 d -Motor overcurrent (+.25 ea) -loss of voltage on the bus 8 ", a se <e4d -low lobe oil pressure . N o W# #4 ta d s u' * * # ' # b' -low suction pressure O'C' m(*dt"go a,,,y"g'.A b ' M I84 GN -SI si3nal os -4 KV bus lockout The 'irrt ' m will cause an idle MFW pump to auto-start (+.5) fi/u M Ned REFERENCE cus ~ r lnsate _and Feedwater Systems', pp 18/19 TPT SD112 'Cond rtr ows o sm a sa 059/000; K4.16(3.1/3.2) ANSWER 2.17 (1.75) 7 p,. g,4 - C #' M A N #'#"'" #* # # EMERG: Instrument Air Compressors (+.5 ear lubrIcatingwat[er-[$",I'./.7St7d # ~ = Turbine Lube Oil Coolers (+.15 ea for any 5) NORMAL: Intake cooling water Back up lube water for cire pumps oil cooling water for primary water deaerator vacuum pump nuclear plant sand filter backwash sanitary plumbing washown hose connections brsW1 ase cwperstera - w a 4 4 e +< wM p..r REFERENCE TPT SD153 ' Fire and Service Water', p 5/6 rPr rac, rb ro - r-e- vo i r', '/u t, W 9 2 ovok 3,g.f g 4 ow e ') )c/, pe 'a-,CA*' M /P t 076/000; K1.19(3.6/3.7)

ja ne89 UNITED STATSS ,.f k NUCLEAR REQULATORY COMMISSION e o nEOsoN N 3 \\ 101 MARIETTA STREET, N.W., SulTE 2000 I ATLANTA, OEoROIA 30323 hf.eWJ.&ESIGNINCLUDINGSAFETYANDEMERGENCYSYSTEMS 2. PAGE 38 ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WM ANSWER 2.18 (1.00) prevent water from flashing on the shell side of the regener.,tive Ht Exensr (caintain RCS pressure in the HL Exchgr with the high temp water) (+.75) cnd prevent damaging RV-203 (LTDN Relief) (+.25) REFERENCE SONP System Descrip, 'CVCS*, pp 9 TPT SD13 'CVCS', pp 17 004/010; K4.03(3.1/3.6) ANSWER 2.19 (1.25) a) 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 035/010; K1.01(4.2/4.5) , p / / ANSWER 2.20 (1.50) f p e) damper trouble annunciator (t d b) Accordian type mechanism 'a fusible leak (+.5) if hot gases pass thru duct, fusible link s (at approx 165 des F) releasin3 the damper a r.d sealin3 the du d +.5) REFERENCE Plan for Requal Cycle IV-1985, ' Appendix R Update' TPT Les i O 003 A1.04(2.7/3.3) l

pa neep gwgTED ST'.TES k NUCLEAR REULATORY COMMISSION 3 o REoioq u

2 I

101 Mt.RtETTA STREET.N.W 8UITE 2900 ATLANTA, GEORGIA 30323 2. PI4WJ.D IGN INCLUDINC GAFETY AND EMERGENCY SYSTEMS PAGE 39 ANSWERS -- TURKEY FOINT 314 -86/02/03-DEAN, WH ANSWER 2.21 (2.00) [uw'I \\' n 30 y o) Tubine is runback at 200%/ min for ;YY seccnds (+.5), stops for.aerra seconds then repeats cycle if condition still exists (+.5) b) The rod position indicating system initiates the runback (+.5) 2: 1:n3 er:te-rau=* is "^!:: ::n::t b' T= bir.: 1:t :t:3: i e r. p l, r, m, m-flou Bo?to'e t tswr r+trun1) )4 Q 1' .5)- o f s: Sia:23 4 RoO bhoPP% b REFERENCE TPT SD127 " Main Turbine Control', rp 18/19 & Fig 11 i)u)G > Toro - T-L - I, JL E ! l 045/000; K4.12(3.3/3.6) l ANSWER 2.22 (1.50) o) RWST (+.25 ea) CVCS Holdup Tanks b) Fuel rack design only allows insertion in specific locations ( +. 5 R 4"St The center-to-center distance ensures <.95 Keff for spent fuele (+.5) (even if unborated water h's used)oes+wq MAT t. ( f $~) 1.%s n. AMt2 covnou A pavin av A o REFERENCE TPT SD41 ' Spent Fuel Pool Cooling, Purification and Ventilation', pp 4-6 75 3 3 thAv t 033/000i K4. ( e ( 3 1/3. 3 ) l

pame:oq UNITED STATES g k NUCLEAR RE ULATORY COMMISSION e o REGeonn I 101 MARIETTA STREET. N.W., SutTE 2900 8 ATLANTA, GEORGIA 30333 3. MSIFtiettNTS AND CONTROLS PAGE 40 ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WM ANSWER 3.01 (1.00) l a REFERENCE Cct, Plant Summary Manual, p. 6 TPT SD5 ' Rod Control System", Fig 13 001/000; K4.03(3.5/3.8) CNSWER 3.02 (1.00) b REFERENCE TPT SD117 'AFW', pp 10 013/0003 K4.04(4.3/4.5) ANSWER 3.03 (1.00) p b REFERENCE , y' Gamma Metrics Neutron Detector' Requal Cycle IV-1985 TPT Lesson P n 01" i K6.01(2.9/3.2) ANSWER 3.04 (1.50) o) V (+.5 ea) b) 2 c) 1 REFERENCE n,'m - r n - s > Fig 11, 12, 13 y b, cr.,a cdr DA S/Cs', TPT SD11 rv Aac 035/010i K4.01(3.6/3.8) l

gasaseg geestga STATSS g%, NUCLEAR REGULATORY COMMISSION RE040N N 2 e 2 \\ 101 MARIETTA STREET, N.W SUITE ~3900 8 ATLANTA. GEORGIA 30333 %,2;oyegNTSANDCONTROLS PAGE 41 ANSWERS -- TURKEY POINT 384 -8t/02/03-DEAN, WH ANSWER 3.05 (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) ANSWER 3.06 (1.50) c) Non-Urgent (+.5 ea) b) Urgent c) Urgent REFERENCE TPT SDS ' Rod Control System', pp 46-48 001/010; K6.05(2.9/3.2) ANSWER 3.07 (1.50) a) False (+.5 ea) b) False c) True REFERENCE TPT SD63 *ESFAS', pp 32,52, FIG 14 006/000; K1.02(4.3/4.6) ANSWER 3.08 (2.00) a) 5 (+.5 ea) b) 4 c) 1 d) 5 (amt d 5 (* (U^' REFERENCE Ty,Sg5 ',R Cgn,ttg stem', FiS 12 001/050; K4.01(3.4/3.8) I L

- "%q UNatto sT:,Tas

,f k NUCLEAR REQULATORY COMMISSION .? n naosoN N f \\* 101 MARetTTA STREET, N.W., sulTE 1900 8 ATLANTA, GEOAGIA 30323 o, / ______litt*__________________ ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WH ANSWER 3.09 (1.50) a) 5 (+.5 ea) b) 6 c) 1 REFERENCE TPT SD13 'CVCS*, pp 65 TPT Requal Cycle IV-1985, ' Heat Tracing', pp 5 004/000; K2.01(2.9/3.1), K2 03(3.3/3.5), K2.07(2.7/3.2) ANSWER 3 10 (1.00) Level III Channel (+.5) failed high (+.5) REFERENCE Westinghouse PWR Systems Manual ' Primary System Control', pp 12-14 TPT SO9 'PZR and Pressure Relief', pp 38-40, 57; DWG 5610-T-D-15 011/000; A2.10(3.4/3.6) ANSWER 3 11 (1.00) 862A (RHR suction from RWST); 863A (RHR recirc to RWST/ Alternate LHSI); 210; RWST Suction Header (+.25 ea) REFERENCE TPT SD21 'ECCS', pp 21 005/000; K4.07(3.2/3.5) relHLb ANSWER 3.12 (1.50) g eg g gg cgaf) [er y oc. c) Loss of voltage sensed by 2 relays on either 4KV bus A or B (+.5) cu tt r b) -Manually reinitiate the SI signal upon the loss of voltage (+.5 ea) -Manually restart SI equipment taking care not to overload diesel REFERENCE TPT SD170 ' Emergency Bus Stripping / Load Sequencing', pp 9, 16-17, ty6 0% t%lO-T*L - I s k n } 064/000; K4.1D/K4.11(3.5/4.0)

l an assu UNITED STATES \\'% NUCLEAR REGULATORY COMMIS810N / Rao@N il I tot uaAlsTTA STREET,N.W., SulTE 2000 I ATLANTA, OEOAQlA 30323 %. ;p#ENTS AND_ CONTROLS PAGE 43 ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WH ANSWER 3 13 (1.25) 3; li 2i 48 10 (+.25 ea) REFERENCE TPT SD4 'Excore NIS', pp 69-70; TS Table 3.5.1 015/020i PWG-7(3.5/4.0) ANSWER 3.14 (1.20) 1. IRH High Flux Reactor Trip (+.4 ea) 2. IRM High Flux Rod Stop 3. PRM High Flux Reactor Trip - Low Setpoint REFERENCE Cat, SD-IPE, p. 19 SONP PLS pp 5 TPT SD4 'Excore NIS', pp 76 015/000; K4.07(3.7/3.8) ANSWER 3.15 (1.80) -PZR hist water level (+.3 ea) -PZR lo pressure -Lo primary 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)

UNITED 8T.'.Ttt /pa aCoq%' 1, NUCLEAR RE2ULATORY COMMISSION P' atoeon n U I 101 MARETTA STREET.C.O., SUITE 2900 ATLANTA, GEORGIA 30323 4SJS114t[NTSANDCONTROLS PAGE 44 ANSWERS -- TURKEY POINT 314 -86/02/03-DEAN, WM ANSWER 3.16 (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) ANSWER 3.17 (1.00) (+.25 eafor 9 4) 1) 862 A and B (RWST to RHR) 2) 864 A and B (RWST isolation) 3) 865 A, B and C (Accumulator Isolation) 4) 866 A and B (SI Hot Les Injection) Pu 4 o43 s)EFERENCE R TRT SD21 'ECCS', pp 36 TPr T 5 L 4 l a.9 006/000; K4.08(3.6/3 7) ANSWER 3.18 (2.00) ,\\ n ~ qp w e c. 4,g, c,c a c) 1-Safety Injection ( +j }y a) 2-Loss of voltage on 4KV buses (f)W IM'F"d bg *# <a. *. 6a4 rfr prep /n/ 3-Lo-Lo level in ANY S/g 4-Loss of BOTH HFW pumps b) 1-solenoid valve in thq control air line to the flow control valve opens (+.4 ea)feraw t/ 2-the electrical turbine overspeed signal to the T & T valve is over-ridden (they will automatically reset) 3-Acw pw%p t+e(% /;u,4I&dnz.~ va /v4 e tel e

REFERENCE I1.to - T-68. u.p 1 4,, 81

'AFW', p TPT SD117 ear n -pr uc d61/000; A4.02(4.5/4.6)

O UNITED ST.i.TES

ga mee k NUCLEAR RE:ULATORY COMMISSION ug n

neosoN N 3 I t oi manierTA 8insE7, N.W., SUITE 3000 8 ATLANTA, O(OnGIA 30323 \\ qopilENTS AND CONTROLS PAGE 45 ANSWERS -- TURKEY PVINT 3&4 -86/02/03-DEAN, WH ANSWER 3.19 (1 50) 1) Hi side of S/U Xfrar energized (+.3 ea) 2) Unit 3 S/U Xfrar has been bus cleared 3) DG Breaker to 4A bus open 4) 4A to 4B bus tie open 5) 4A bus not locked out REFERENCE TPT SD170 ' Emergency Bus Stripping / Load Sequencing *, pp 6-7 000/055; EA2.03(3.9/4.7) ANSWER 3.20 (1.75) a) Cabinet has the capacity to support up to 6 stationary gripper coils simultaneously (+.5). So with 2 groups or more, would overload / heat the cabinet (+.5). ff(su^ k cud / ceud rok h dred b) 125 VDC-Latching Rods 70 VDC-Holding Rods (+.5 for reasons, +.25 for correctly associating voltages) REFERENCE TPT SD5 ' Rod Control System', p 42- ?.,J itch Ma n nA, M Lv+ w M J co s trol,p@ L Z,4 3 001/050; PWG-1(3.6/4.1) ) ANSWER 3.21 (1.00) s e p a r a tion (+ 0 7)

1) you could parallel MCCs on different units (eliminate unit (W

2) the MCCs could not be in syne if paralleled causing cirevlatin3 currents a:id create overload conditions

( + 5r ) b'N REFERENCE TPT Lesson Plan 20-OL, APP B ' Replacement of 120 VAC Inverters *, pp 5 Requal Cycle IV-19H5 Ter W4 s 6 te T.s -/JY1 062/000; A3.04(2.7/2.9)

UNITED STATts 4',ggneeug% NUCLEAR P.E2ULATORY COMZl8SION ,=g REGeow n n I 101 MARIETT A GTREET,N.W SUITE 2900 ATLANTA, GEORGIA 30323 3. RSTSW [NTS AND CONTROLS PAGE 46 ANSWERS -- TURKEY P0 INT 3&4 -86/02/03-DEAN, WH ANSWER 3.22 (1.00) Direct immersion (no weld) decreases the time it takes the RTD to respond to a temperature change. REFERENCE VCS, IC-6, Temperature Indication System, p. 14 TPT SD7 'RCS', pp 25 016/000; K1.01(3.4/3.4) )

i UNITED STATES ,# gha aerog% NUCLEAR REGULATORY COMMISSION 3 o RE0lON H 3 101 MARIETTA STREET N.W..SulTE 2900 8 ATLANT A,0 EOR 0!A 30333 QG5paJRES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 47 --- EX5i5E55iEXE 55ETE5E------------------------ ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WM ANSWER 4.01 (1.00) o REFERENCE TPT E0P-FR-S.1 001/010; A2.08(4.4/4.6) ANSWER 4.02 (1.00) a REFERENCE TPT ONOP 2608.1, pp 1/2 000/024; PWG-10(4.1/4.4) ANSWER 4.03 (1.00) d REFERENCE Westinghouse background info for TPT E0Ps, 'RCP Trip / Restart *, pp 49/50 000/074; EK3.07(4.0/4.4) ANSWER 4.04 (1.00) REFERENCE 10CFR20.5 PWG-15: Radeon Knowledge (3.4/3.9) l l l l

g>m88e9 U0NTED STATES 9 k NUCLEAR REQULATORY COMMISSION At040N 11 .? e 5 I 101 MAAltTTA STREET.N.W SulTE 2900 a ATLANTA, GEOAGtA 30323 4. P'ROCEpelh S - NORMAle ABNORMAL, EMERGENCY AND PAGE 40 ~~~~ 565 E 556AE~6 iRUE~~~~~~~~~~~~~~~~~~~~~~~~ R (.NSWERS -- TURKEY POINT 3a4 -86/02/03-DEAN, WH ANSWER 4.05 (1.50) o) False (+.5 ea) b) True c) False REFERENCE Wastinghouse User's Guide for TPT E0Psr pp 5-12 ANSWER (2.00) (0.5) 49 v A^^ "~' o. -E,__. 7 <~ i 6. Fast (0.5) c. at (0.5) d. Governor control (0.5) REFERENCE Cetr OP/0/B/6350/11, enclosure 4.4a pp. 1-2 TPT OP 4304.1, pp 6/7,SP,H prv'/ bh,~pr hl/, pp 7/8 0 064/000i A2.03 and A2.09 (3.1/3 3) 4 (1.00) roweledo o. 07 ANSWER r 3; Turbinei Boron; 75% (+.25 ea) REFERENCE TPT ONOP 1608.1, pp 9/10; TPT TS B3.3.2 000/003; EK3.04(3.8/4.1) ANSWER 4.08 (1.00) a) 10EE(-8); 1 (+.25 ea) b) i c) 1000 REFERENCE TPT OP-0202.2, pp 2/3, 18

o [f',># 884 4 UNITED STA78t3 1 NUCLEAR REQULATORY COMMISSION o RE060N il l 101 MARIETTA STREET. N.W., SUITE 2900 F ATLANTA, GEORGIA 30333 \\,.. f 4'/FS - NORMAL, ABNORMAL, EMERGENCY AND PAGE 49 /. ~~~~~~~~~~~~~~~~~~~~~~~ Rd656L6656A6~66OTR6L ~~~~ ANSWERS -- TURKEY POINT 384 -86/02/03-DEAN, WH 001/050; PNG-7(3.6/4.1) ANSWER 4.09 (1.50) .3 1) Tavg/ Tref Deviation Alarm ( +.M e a ) [4//tho< O G56J#f '" 2) Overpower Rod StoP Sr M d 4W/A 3) RCS High Delta T 4) OP/0T Delta T Rod Stop 5) RCS High/ Low Tavs 6) Rod Bank D Low Limit Alarm REFERENCE TPT Requal Cycle II Lesson Plan-1985; NRC Generic Letter 85-05 of 1/31/85; TPT ONOP 2608.1/2; TPT SDS ' Rod Control", pp 24 & SD7 'RCS', pp 69 004/020; PWG-10(4.3/4.5) ANSWER 4.10 (1.50) 1) CNTH Air Particulate Monitor (Any 5 for +.3 ea) 2) CNTH Radioactive Gas Monitor 3) Component Coolin3 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 2 000/028; EA1.06(3.3/3.6) ANSWER 4.11 (2.00) 1) Place SJAE in srve (+.33 ea) 2) Place hoggin3 jet in srve 3) Reduce turbine load to keep vacuum above trip point of 20' Causes: Loss of sealing steam; air leakasei circ water supply header high tempi loss of cire water supplyi loss of circ water box primei SJAE failurei SJAE loss,of loop seali Open steam traps;

1. 0 ) (a ddhEJ (*'v "u w w1 G:c Q+*4) h

(+.25 ea up to REFERENCE TPT ONOP 6108.2, pp 2 O m P -o *'4, PP V [Myroceduct)

e /g*"C&g USNTED STATRS ,g NUCLEAR REEULATORY COMMISSION 3 e naceok u 2 I 101 MARIETTA STREET N.W., SutTE 2000 ATLANTA, GEORGIA 30323 o, hRGCgpalliES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 50 4. ~~~~~~~~~~~~~~~~~~~~~~~~ ~~~~ d656L665CdL"66YTR6L R ~ ANSWERS -- TURKEY P0 INT 3&4 -86/02/03-DEAN, WM 000/051; PWG-11(3.7/3.7) [Z o f 5) ANSWER 4.12 (1.00) Centainment pressure (+.35) > 4 psis (+.15) 10EE5 R/Hr (+.15) Containment radiation,F( +. 3 5 )) >3 Can T'. % P O st) > t90 (.+. r t ,, u REFERENCE Wastinghouse backaround info for TPT E0Psr " Instrumentation Accuracy'epp 11 se SerTIST $1590 Y TDti dDP G f MAAW/P /# 9 022/000; K3.02(3.0/3.3) p,q % pg) ANSWER 4.13 (1.50) 1) Manually trip reactor (or open trip breakers locally) ('.5 ..) _2,) Manually trip turbyne pg. g g g,j u,u,, gg4 pf 1;3 :: g :: - c., y () Evacuate CR and,v v#c-- o o. u.. r e r...cr :; r. ;..mv pcr ,7::::: f rue 6at opsm)o: ckad (4.S) REFERENCE ( TPT 0-ONOP-103, pp 5 % W ckup cu k reveJ.- ofh/M/pr 000/068; PWG-11(4.5/4.5) Q ANSWER 4.14 (1.00) 1) RCP Thermal barrier CCW outlet (FCV-626) (+.25 ea valve) 2) RCP Bearing CCW outlet (MOV-730) 3) RCP CCW Supply isolations (MOV 716A and 716B) REFERENCE TPT SD63 'Rx Protection and Safeguards Actuation Systems' pp 33 103/0005 A2.03(3.5/3.8) ANSWER 4.15 (1.50) 1) PZR PORVs Closed (+.5 ea) 2) LTDN Isolation Valves Closed 3) Excess LTDN Isolation Valves Closed i l l

'l UNITE 3 STATts /jh* " Coq %, NUCLEAR RE ULATORY COMZISSION a e o REGeon n 3 101 MARIETTA STREET, N.W., SulTE 2000 ATLANTA, GEORGIA 30323 kROG5pdES-NORMAL, ABNORMAL, EMERGENCY AND PAGE 51 4. --- sA5i5E55i5AE 55siE5E------------------------ ANSWERS -- TURXEY POINT 3&4 -86/02/03-DEAN, WM REFERENCE TPT E0P-ECA-0.0, pp 3 000/056; PWG-11(4.5/4.6) ANSWER 4.16 (1.50) Core Exit T/C (+.5 ea) T-Hot RCS Subcooling REFERENCE SON ES-0." pp 6 TPT ES-0.<. pp 5 EPE-074; EA1.02 (3.9/4.2) ANSWER 4.17 (3.00) o) -SI Annunciator ON (+.3 es response) -SI pumps running -RHR pumps running -EDGs running b) -CNTNT Purge / Supply fans OFF -Purge Valves CLOSED -Instrument Air Bleed Valves CLOSED -Verify Control Room Ventilation Isolation c) -AFW steam supply MOVs OPEN -AFW Flow Regulator Valves OPEN REFERENCE TPT E-0, pp 4/5 000/007; PWG-11(4.4/4.5)

gna mag geelTED STATES g g k NUCLE /.R RE2ULATORY COMMISSION i .E l. e 5 \\ 101 Me AltTTA 8TREET.N.W., SulTE 2000 t ATLANTA, OEORGIA 30323 4. P'RSCgp417t S - NORM AL, ABNORMAL, EMERGENCY AND PAGE 52 ~~~~ d656L665CIE~6 NTR E'~~~~~~~~~~~~~~~~~~~~~~~ R ANSWERS -- TURKEY POINT 3&4 -86/02/03-DEAN, WH g g # alu MM f ANSWER 4.18 (1.50) -Take Local / Remote switch to Local for Group A (+.4) -Close breaker using local pushbutton control (+.4) -Both of these controls are in the NORTH Electrical Penetration Room (+.2) -Indications of being in Local are:-Both Red and Green Breaker Lights ~out

REMOTE-LOCAL CNTRL SW IN LOCAL' n.f w - 7 /,*g&.J annunciator

(+.5) e scHl. REFERENCE ') M ** /Cy lu/ges Av b Vff r# Q4'"1 WO Who Relief', pp 21/22, Fig 13

- f*'* V6 TPT SD9 'PZR P

fW 1)of ldh 6 - Pf!knd Er ssurk,tWV 6Lh a n W fo*#e/ > f A q 005 O h10/000i A2.01(3.3/3.6) yCPerg L4I' % 5 ANSWER 4.19 (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 E0Ps, 'RCP Trip / Restart' 000/009; EK3.23(4.2/4.3) ANSWER 4.20 \\(2.00) (pe'4 3 a) Lettering of sub-tasks (+.5) b) Return to next step or sub-step on the left side (+.5) c) Monitor all remaining trees for RED terminus (+.5) and if not encountered, suspend any ORP and perform the applicable FRP (+.5) REFERENCE Westinghouse User's Guide for TPT E0Ps, pp 3-11 ANSWER 4.21 (1.00) 3, 1, 2 ( .5 for each swap to put in correct order) REFERENCE Westinghouse background info for TPT E0Ps ' Foldout Page Items *, pp 1

Pcge 1 NRC Exam Secti:n 1 ( QUEtrr10M 201.03: (SRO 5.03, SRO Requal 5.02) '~ 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 ins' ead a 2%/ min ramp was used? t

a. The renon 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 menon dip for the 2%/ min. ramp would occur sooner and the magnitude of the dip would be larger.

d. The zenon dip for the 2%/ min, ramp would occur later and the magnitude of the dip would be larger.

V

RESPONSE

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. I

REFERENCE:

Westinghouse Simulator Training Book - Reactor Theory and Core Physics Page I-5.76. 1 e RI-5/dj _ _ _ - = _.. _ _.

~~~ 07/06/86 Pcg3 2 i NRC Exam Secti:n 1 QUESTION RO 1.13: (RO Requal 1.98) t o What are the two primary factors that provide the driving mechanism for Natural Circulation flow?

RESPONSE

We request that "a heat source and a heat sink" with "a AT' also be considered as an acceptable answer. The heat source is implied while the heat sink and AT are speelfleally addressed in our training materials.

REFERENCE:

Westinghouse Text " Mitigating Core Damage" Chapter 1, pg 11 and 12. i N.. eRI-5/dj +

" '~ 02/0G/86 Pcg2 3 NRC Exam Secti:n 1 [ QUEirrlON RO 1.14: (SRO 5.14) List three significant heat transfer advantages of a counter flow heat exchanger over a parallel flow heat exchanger.

RESPONSE

We request that "a counter flow heat exchanger la more efficient..." be an acceptable answer.

REFERENCE:

Westinghouse Text " Thermal-Hydraulle Principles and Applications to the PWR" pages 5-39 and 5-40. l I l i s eR1-5/dj

02/06/ss Pcga 4 i NRC Exam Section 1 i:f : ; QUErrION RO 1.16: (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

i' 1-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 graph 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 n'.N EOL. To determine the differential boron worth from this data, the following sequence would s be' followed. 1) Determine the integral worth for a given borgn concentration b

2)

Determine the correction factor for the total Xenon and Samarium in the reactor i for the desired conditions. ~ /"- 3) Divide the integral worth by the correction factor to etermine the corrected l 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 t,oron 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: . R 1 -5/dj l

= 02/06/86 Pcg2 6 NRC Exam Section I g,'.. ; i.-].[- 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? h

RESPONSE

W 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 9 e y e

y 9

e i ~. s .R1-5/dj

02/06/86 I Pcg2 7 e i NRC Exam Section 1 b QUESTION RO 1.21: (BRO 5.23) Using the attached steam tables, what is the amount of primary subcooling at the core exist if the pressurizer is at 2235 psig and Tavg is 575 degrees? (assume normal operating conditions)

RESPONSE

(..n. We request that a 500F i 20 value for subcooling be an acceptable answers If student assumes: AT = 560F and Tsat for 2235 = 653oF; then 575 + 56/2 = 6030F = Th ..subcooled margin = 653oF - 6030F = 500F; the correct answer can thus be 500F 12o. 4

REFERENCE:

y s. SD No. 7, page 68 b .Y .f~ o t t .RI-5/dj

02/06/86 ~ Pcga 8 NRC Exam Section 1 QUESTION RO 1.22: (RO Requal 1.13) 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

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 + p / [-p seentrached I s. 3 I: .~ 6 . R I-5/dj m -- - -,,,, - + -, - --p -w --o

I

c. no w Page1 NRC Exem Sestion 2 QUES'110N RO 2.391

%g.e, Whleh of the following is NOT a source of water to the PRT? a. Reactor Vessel Flange Leakoff Detector Drain b. Letdown Relief Valve RV-203 c. R'CP Seal Water Return Line Relief Valve RV-382 d. Safety injection Test Line Relief Valve RV-859 e. Accumulator Discharge Line Drain RESPONBE We request you also accept answer "a* as a correct answer as reactor vessel flange leakoff drains to the reactor coolant drain tank, not the PRT.

REFERENCE:

, p.. TP-5610-T-E-4501, sheet 1 of I j sv t s. t 4 y } l

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Page3 i l NRC Exam Se::ti:n 3 QUESTION RO 1.03: (RO Requal2.03) Whleh of the statements below regarding Unit 3 APW pump steam supply valves on a loss of the SA 4EV bus voltage is correett a. Steam supply valves from all 3 8/Gs open. b. Steam supply valves from A and B 8/Os open. I e. Steam supply valves from B and C 8/Gs open. d. Steam supply valves from A C S/Os opea. t

- m...w e gh*; ae..*

, gY e. lNo steam supply valves wBl apen es It takes loss 'of voltage on both.4EV bus f SA and SB to cause the valves to.'open. 2-RESPONSES We roguest you accept answer "If as the correct answer, as loss of voltage on 4KV bus 'A' opens steam supply MOVs from "A" and "B" steam generators. This feature is oommon to units 3 and 4.

{9)

REFERENCE:

'g r t-TP-5610-T-LI, Sheet 15 p ,? f r ... _ _. _ _ _ _ _. _ ~. t

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~~~ 02/06/86 Page3 NRC Raam Sectisn 2 (/, QUERI10N RO 8.08: (RO Requal1.04) Match the RC8 penetrations in Column A with the appropriate RC8 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 L

1) Loop A coldleg

..,. 1) Loop A. hot leg,>,. b. Par Burge Line e. Alternate Charging 3) Loop A'latermediateleg d. Par Spray Line 4) ' Loop B latermediate leg e. RHR Suction 5) Loop B hot leg 8) Loop C cold leg ' F) Loop C hot leg Ru8pousa h MC We request you accept answer #7 for part e of this question as alternate charging is assoelated with Loop C Hot Leg. s We request you accept answers #2 or #7 for part e of this question as units 3 and 4 ,Miffer in this respet. Answer #7 or C Hot Leg applies to unit 3 and answer #2 or Loop ' A Hot Leg applies et unit 4. . REFERENCPA Part C ~ TP-5610-T-#-4501 Part E TP-5610-T-#-4510

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Pega4 NRC Exam 8esti::n 2 (.[. QUE8110N RO 1.16: o List the 8 MFW pump trips and indleate whleh of the trips will cause an idle MFW pump to auto-start. (Setpoints not required) REDONER: I We request you accept two additional answers with respect to steam generator feed pump trips as these pumps also trly on bl/bl steam generator level and trip of an assoelated condensate pump. r Additionally, with respect to the auto-start feature, the idle steam generator feed plump auto-starts on overcurrent trip of the running feed pump and low lube oli pressure on the running feed pumps the first and third answers listed on the answer k.ey.

REFERENCE:

i ? TP-5819-T-L1, sheet 25A l 4 i 2 i. s. .'I e 6 + I I

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~ 02/06/SC Page5 NRC Exam Section 2 (ifj QUESTION RO 1.17: a) List the 2 Emergency loads supplied by the service water system. b) List 5 of the 6 normalloads supplied by the service water system.

RESPONSE

1 In addition to emergency backup to the instrument air eompressors and turbine tube oil 'E ecolors, we request you accept the following answers

s

' Emersoney M Charrint Pumos

REFERENCE:

TP-5810-T-E-4075, sheet 1 of 1 and ONOP 3108.1, Loss of CCW Flow [ ] Auxillark Feedwater Pumns I

REFERENCE:

TP-5818-T-E-4082, sheet 8 of 7 ( Condensate Storare Tanks

REFERENCE:

8D-117, page 19 and ONOP 7308.1, Malfunetton of the Auxillary .j s. ~ f Feedwater System, page 15. ~ t Fire Protection ,kEFERENCE: TP-5610-T-E-4072, sheet 1 of 3 f Mormal Loads Breathinz Air Comoresso.rf.

REFERENCE:

OP-15650, page 3, T.reathing Air System Operating Instructions Feeds Water Treatment P1gnt.

REFERENCE:

TP-5610-T-E-4530, sheet 4 of 4 l ~ 'RI-5/dsp

02/06/86 Peg 2 6 NRC Exam Section 2 QUERI10N RO 2.20: (RO Regual S.14, SRO 8.19, BRO Regual 4.12) a) What indication does a control room operator have that a fire damper has actuatedt b) Explain in detail the design features which allow a fire damper to auto close when ';'Q required. n:.:;i l

RESPONSE

1 We request this question be deleted for the following reasom 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 was presented to licensed operators as "Look Ahead" to inform i g them of the appendix "R" modifications in progress. e presentation given to the {j f Group I students did not cover the fire dampers) Tiai'e instruction given was not detailed because the modifications were still in progresa. More detailed training will be provEl'ed when the appendix "R" modifications are completed and ready for turnover. g ? ^ marnRnNca: 8 y 0 ~, Lesson 82-OL Appendix A page 4 titled Appendix "R" Mods (Presented to Group X Students) @~. 3'

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7 h Al[ PCE* 7 NRC Exam Sesti:n 2 1.~ fj b QUE8110N RO 1.21: SRO 8.20, BRO Requal 4.13) o p a) Describe the runback process that occurs with the Main Turbine when the OT Delta T setpoint is exceededt b) If the Power Range " ROD DROP AUTO TURBINE RUNBACE" la bypassed, what conditions must exist and what system will initiate a turbine runback? RESPONEBs ~' > ";ijf.2. ?:l.:f." t ".' ' ~ ?. Part A We request the NRC answer key be corrected to read as follows:

a)

Turbine is runback at 200%/ min (Unit 315816 inta.) for 1.5 sec., stops for 30 see. then repeats eyele if condition still exists." i.l 7 - Part B: We request that the portion of the answer " what conditions must exist..." be deleted. s. The condition ". Reactor' Power is 70% ~as sensed by turbine 1st stage impulse i pressure." is incorrect for the situation addressed in the question. ( t

marnanmes

t-Part A

Dwg 5610-T-L-1, ah. 21 Part B Dwg 5610-T-L-1, sh 21 l
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02/06/86 Pcg2 8 NRC Exam Section 2 h QUESHON RO 2.22: (SRO S.21) a) What are the 2 sources of Borated water available for the Spent Fuel Pool? b) What 2 design features of the spent fuel racks ensure criticality does not occur in the Spent Fuel Pool? 't-

RESPONSE

r6 Please include as an acceptable answer to part B that the racks contain a neutron absorbing materlaL .I f

REFERENCE:

i i T.S. B-3.17-1 para. 2 ) I ll _} t-4 <T. l g..' 1 e f

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02/06/86 Paga 1 NRC Exam Section 3 QUESTION RO 3.03: (RO Requal 3.02, SRO S.95, SRO Requal 6.84) Which statement below correctly describes operation of the GAMMA-METRICS Neutron Flux Monitor in gamma flux fields between 10,000 and 1,000,000 R/hr (ie. high radiation fields), s. The monitor is not designed to operate in such high level radiation fields. K i M. Y 'P b. The monitor will operate satisfactorily in these radiation levels, but an g adjustment should be made to discriminate against the higher gamma flux.. o c. The monitor will operate satisfactorily, but the output signal from the ij detector will not increase linearly due to lack of voltage saturation in the detector. i d. The monitor is designed to operate as well in this high a level gamma flux as l y it does in much lower radiation fields. 1

RESPONSE

S-We roquest this question be deleted for the following reasons, l ij The Gamma-Metrics Monitor is a component of the larger safe ahutdown syn..em. The . Safe Shutdown System has not yet been turn-d over for plant use. A partial turnover of the. Gamma Metrics Monitor _was. performed in_1985. The Gamma-Metrics Monitor was ] ,Q functional at that time but was to be used for indication only. Procedure changes and ~ i 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 l operators as "look ahead" to inform them of the new instrument in their control room. However, the instruction given was not detailed because of the limited purpose of the Gamma-Metrics Monitor at that time. 'R1-5/dsp

N 1 Page2 NRC Exam Bestion 3 QUBrrlON R'O 3.64: (RO Requal3.03) h..i o Indleate whether there are 1, 2 or 3 8ELECTABLE detector inputs for each of the following parameters utilised by the 8/G Water Level Control System. a) S/G Level b) Feed Flow i e) Steam Pressure ] 2 t RESPONSES -} 3 We are requesting that you accept 2 (two) as the' number of selectable S/G level 1 detector inputs to the S/G level 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 mhannel was added. i,

REFERENCE:

Plant drawing 5819-TD-17 1 (, l. y I l l I l l

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es7mmw. Page 3 NRC Ex m Sestion 3 QUESTION RO 3.08: - (RO Requal 3.08) v. Match the interlock descriptions in Column A with the appropriate logie 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 4143% power 1. 1/2 b) Overtemperature Delta T rod stop 3. 1/3 e) Intermediate Range High Flux 3. 1/3 ] d) Power Range Rod Drop 4. 2/3 5. 1/4 s. 2/4 ~ T. , 3/4 g 2 REEFONSE: i 0 W We are requesting that "5 and/or 8" be accepted as correct ' answers to part "D" to this { questloa. 'lh Power Range Rod Drop logie is currently different for Units 3 and 4. j J 1-The correct answer for Unit 3 is "5"(1/4). e The correct answer for Unit 4 is "$" (1/4). i, e y g

- As the Unit number was not speelfled in the question, either or both of the above should be acceptable. -

~

REFERENCE:

Plant drawing 5610-LI-1, sheet 17 See Note 1

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Feo 4 3 a NRC Exam Sectt:n 3 quBrnou ao s.13: (no moqual s.es, Sao s.te) b. a) Olve the location and the number of UV relays that must be energized to initiate bus stripping on a Loss of off-8ite Power. b) To ensure needed vital equipment starts on a Loss of off-81te Power following an SI that has been RESET, the operator can perform what two actions? j RESPONSES

l*

' E: ].)[,,0.'.@O 'i- {; We request you accept the following additloast answers for' Part.a of this question. .i 2 Loss of voltage relays per bus (physically located inside the sequencer cabinets) c } Under voltage relays on the associateId 4:8V lead senters. 4, s .i p$ 8D-170 page 6 1 ..~ 'bwg. 5810-T-L-1 sh.13 ~ 5 'RI-5/dsp

s.so u n/ 6D s Pago 5 S NRC Exam Section 3 (.,5 QualmON RO 3.17: (BRO 4.14, SRO Requal 4.99) r g List the 4 sets of ECC8 related valves required to mitigate a 14CA whleh have their control power breakers racked out during eritleal operations.

RESPONSE

We are requesting you necept valves 883 A & B as an additions 1 oorrect answer. These valves are required to be controHed by T.S. 3.4.1.r f on page 3.4.1. I I 1 0, s. 3 .Y ~ s l 6 I

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t:2/0GRI.j Pcg3 6 i NRC Exam 8esti:n 3 (tUE8'I10N RO 3.18: a) List all the Auto-8 tart signals for the AFW system. b) Aside from opening the AFW Turbine Steam Supply valves, what other 2 actions associated with the AFW system occur upon an initiating signal?

RESPONSE

F.8ELA ~' We request that the following be socepted as correct answers. ~ Bus Stripplag, & Loss of Vottage on either 4EV bus A or B (initiates' bus stripping) g Under Voltage on 480V Load Centers A, B, C, or D (initiates bus stripping) Also request that answer #4 be' modified to accept " Loss of all running 8/G FW pumps". Pm b ~ i We requ.est that the following be accepted as an additional correct answer g "AFW Pump 5 team Line Inlet Drain Valves Close". This coeurs automatleally on AFW pump start at 150 psig increasing steam supply prcasure.

REFERENCE:

~ Parta I5610-T-LI, sheet 15 Note 1 and logie in upper left Partb SD 117. nare 11 and 5610-T-LI, sheet 15 i

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- ~.

cJ7SA'M / W Pego 7 NRC Exam Beettrn 3 QUE8'110N RO 3.20: SROS.22) e, i 1 a) What consequences could be expected in the Rod Control System's DC Hold Cabinet if 2 or more groups of rod drive mechanisms were place d on Imld power (excluding Control Bank D rods)? Explain your reasoning? b) Why is there both a 125 VDC and a 70 VDC power supply in the DC Hold Cabinet? REEFONEEs w..."_,.hy. ..;:q

.-J.

+ Please accept as an additional correct answer to part As } 1 " Fuses will blow and/or rods will drop." The embinet has the capaelty to support a maximum of 8 eontrol rods {1-1/2 banks). To 4 place two or more groups on it will result la overloading the supply (M-G sets) whleh causes the fuses to blow and at least two rods to drop. ..:p

REFERENCE:

~

1. West Tech Manual Full Length Rod Control Vol.1 i

l y l 1

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Page 8 a NaC Exam Becti:n s quasTion no s.21 (no moqual 3.13, sao s.2s, sao Requal s.14 , 63 The Alternate Source Transfer Switches associated with the recently installed 120 VAC inverters have key locks to prevent 2 switches of the same channel being selected to ALTERNATE at the same time. What are the purposes behind this admin 8 trative key J control?

RESPONSE

b .i.,,, t;. ;.,

., ;.; c y

Please accept as an additional sorrect answer for part l et this questloat 2 "You could parallel the CVT's on different units." 4 b

REFERENCE:

-4 t SS18-T-E-1592 g 0 s, i t n i; i s

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f WJ/bO/WO Page 1 ) NRC Ex0m Section 4 g:- quElmON RO 4.96: (RO Requal 4.94) For the following paragraph, choose the correct words from the options given after the paragraph that oorrectly 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 synebro pointer as it moves slowly la the h direction and eloslag the generator breaker when the polat is 2 the vertleal positloa. De power (MW) output of the y, generator is then raised by agusting the 6 ~ Chose from the followings i ~ lower than / equal t'o / higher thea f s. P1 b. slow / fast i e. 5 minutes to / at / 5 minutes after } d. governor control / voltage regulator / stator cooling i O REEPONEE: ~ l ' ' We request that either " equal R" gt " higher than" be accepted as correct answers for / part "A" of this question. ~ { i $Ldjusting incoming voltage alightly higher than dmning voltage has been stressed in ? operator training to ensure a lagging power factor after breaker closure. ~~ ~ t

REFERENCE:

SD 137, page'44 i

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dure, adjust the DG (incoming) voltage slightly higher h than the 4 KV bus (running) voltage to assure a lagging power factor. During operation, paralleled to the system, b adjustment will not normally be necessary. Upon transfer of the Local-Normal Control Switch to Normal, the Normal light on the engine control panel must come on. Upon a manual or automatic unmentary start signal, the engine will start and accelerate to rated speed and voltage automatically. Engine starting sequence in the normal mode is similar to the starting sequence in the local mode described previously., f Automatic Shutdown Automatic shutdown in the normal mode is the same as described in j the automatic shutdown section in the local mode. ? Itanual Stop z .l The procedure for stopping the engine generator set is very much p the same as is in the local mode, except that.in the normal mode both the local and remote pushbuttons are available for manual stopping. I '~ \\- HAISLIlm REACTIVL. LORD OR 11E SIESEL EIERATM ~ Before paralleling the diesel with its bus,"the Operator should { ensure that the nachine is ^up to speed and indicating 60 Hz and I ,f 4160 volts output. Voltages should 'be matched for all phases 9 ~ hNI ? T51s ensures a lagging power factor after breaker closure. In ' eparation for breaker closure, adjust engine speed so that the i p, sinter on the synchroscope is rotating slowly,in the fast direction. This condition means that the incoming machine (diesel generator) is spinning faster than needed to produce 60 Hz. Therefore, it will slow to synchronous speed the instant the breaker is closed. The extra torque (no longer needed for the faster rotation), will be used to pick up electrical load. When l the pointer on the synchroscope indicates 12 o' clock, the two voltages are in phase. There is no potentici difference across I s ~ SD137-Rev.0-44 i --.,_------------.-.---._-..-nn.

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Page2 NRC Exan Secti:n 4 QUESTION RO 4.07: 4.2 Fill in the blanks for the following statement regarding immediate settons on a Rod Drop acendent. A Reactor Trip is required if or more rods have dropped. Assume rods in automatic initially and Reactor does not trip. Initially, T wg is. contro!!ed using the and when temperature is stabilised, h ased to control Tavg. 1 Techeleal Speelfloations allow operation with a 4W rod for over 8 hours as { PI long as power is reduced to < to ensure design margin to eore limits are }l malatained. f r1 l

RESPONSE

( i 5 We request that "controt rods" be considered an additional correct answer for the second I blank. During the exam the examinees were instructed to assume sontrol rods in h automatie. Initial Tave control would include automatie control rod insertion, until rods 3 were piaced in manual by the operator, according to our procedures. i,. D c u

REFERENCE:

1 s.

h.,

ONOP 1808.1, page 9 f Sect. 5.4.2 - Automatic Actions j y + 1 'l

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D2/06/86 i Page3 NRC Exam Section 4 ! g.. J - QUESTION RO 4.09: t'RO f.98, SRO Regual T.05) e List 5 possible alarms (setpo!nts not required) on the Main Control Board that would be indleations that an inadvertent dilution were occurring while the Unit was at power. (Assume Rod Control is in MANUAL, NO Rx Trip occurs an NO operator actions are taken to mitigate the dilution). REEPONSE: ?

f Due the lack of speelfle information with regard to initial condi,tions, this question lends f

~ j. itself to many possible answers. 1 4 We submit the following list of additional alarms which' may be Indicative of an inadvertent dilution and request they be included in the answer keys - ~5; 2l ?e 1 s. 1, '1) g dl t i 'RI-5/dsp

0;/06/86 Pcg2 4 l NRC Ex :m 8esti::n 4 [ 4 ANNUNCIATOR EXPLANATION G 5/1, A$>5% Max Pwr 90% As inadvertent dilution progresses, core axial flux will shift toward bottom of core by increase Tavg in the top of core. With axial flux already near the negative setpoint. This alar m may warn the operator of an inadvertent dilution. G III ChargingPump LoSpeed As Pzr level increases, due to RC5 expansion, chg pump speed will decrease. At 16% of full speed the chg. pump low speed alarm would provide additional warning to operaton. j ~ \\ B-5/1,Interm. RNG. High Flux Level If at low power (above P-10) positive reactivity due to Rod Withdrawal 5 top dilution causes Tavg & Rx power to increase when 1/2 IR inst's reach 20%, annun. B-5/1 will alarm to warn operators of the problem. A1W2 Pressurizer Control Hi/ Low The inadvertent dilution adds positive reactivity causing Level Tavg to increase er d causes RCS to expand. At 5% deviation from actual level verses setpoint level, the annu nciator will afarm to warnthe operator. A3/5 Demineralizer Flow Diverted Due to charging pump speed and thus flow being reduced ,h,, HighTemp (because for Tavg driving PZR ievel high) the Regen heat exchange cooling flow is reduced thus causing letdown temperature to go high. This in effect overloads the Non Regen heat exchanger and TCV-143 will go to divert. This will overload the Non Regen Hx because the CCW valve which controls temperature is isolated & bypassed so the Hx has only a limited amount of cooling water. B 3/6 Overtemp AT As dilution progresses Tavg increases thus lowering our OTAT setpoint. Before we get the Rx trip we get an alarm and a turbine runback / rod stop A 2/5 Rx Coolant Makeup Boric Acid This alarm indicates whenever boric acid or makeup water Flow Deviation to blender station is not within spec as we have set them. A 2/6 Rx Coolant MakeupWater Flow Deviation B 6/1 NIS Power Range Single Alarms when any one NIS channel (N41,42,43,44) indicates Channel High Range Alert above 108% power. This could come in as a result of a dilution accident if Tavg increased enough to allow the NIS detectors to see a higher flux. P 6/3 NIS Power Range Overpower When 1/4 power range instruments reads > 103% power Rod Withdrawal 5 top this alarm comes in. This could come in as a result of Tavg being high and allowing NIS to see a higher flux.

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v4eucine Page5 ,? NRC Exam Bestion 4 QUESTION RO 4.11: a) After trying to identify the cause of a SLOW decrease in the Main Condenser vacuum, what are the remaining three immediate mettons? b) List four possible causes of a loss of Main condenser vacuum.__,_ RESPON58: We request that the following be considered additional correct answers for part '8" of this question: { 1 Turbine expansion joint boot seal not filled with wate,r. 1 Turbine seal stop drala line fa!!ure. 1 8/G FW Pump Seal Water Collection Tank drain to condenser failure. Cire. Water intake Screens blockage

REFERENCE:

-ONOP-014, page 5 1
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s N 02/06/86 ~ kf Page8 -[ NRC Exam Section 4 /g QUEtr10N RO 4.12: (RO Requal 4.07, SED 7.99, SRO Requal 7.96) List the two conditions (including setpoints) which determine Adverse Containment conditions. i RESPONER: We request that 180'F be considered an aseeptable answer. g -f a ;_ - The background information for PTN's SOPS were based, upon the standardised V Westinghouse Owners Oroup bases. However, as plant speelfle data was considered it ,[ was determined that the most limiting instrument would require that adverse ? 1 eontainment conditions be either 180'F or 1.3 K 105.R/hr. !e

REFERENCE:

~ b 1v Wc11aghouse setpoint study for EOP Generation, pg.187 0 4 y t e 1 U

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02/06/86 f Pcge 7 NRC Exam Section 4 QUErrION RO 4.13: (R0 Requal 4.98) What are the Unit 3 RCO's immediate actions if the word is passed " Fire in the control room, shift personnel report to assigned control room evacuation stations"?

RESPONSE

We request that the answer for this question be changed to read: 1. Trip the reactor 2. Trip the turbine 3. Assist the PS-N 4. Evacuate control room - report to turbine operator shack I UEFERENCE:

-ONOP-103, page 5, section 4.4

.t, 'it? s [ 1 6

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02/06/86 Pcga 8 NRC Exam Section 4 QUESTION RO 4.18: (BRO 7.20) A failure in the Unit 4 PZR level control circuitry results in level controllers LC 459C & i 460C sending false Lo-Lo level signals (<14%) to the PZR heater controllers. Describe what must be done to the Group A Backup Heaters in order to restore them to operation with the false level signal in place. Indicate locations of any controls operated and any indications in the control room that any remote controls are operated.

RESPONSE

We request that you accept as an additional correct answer te this question the' following: [ { Proceed to Rack 46 (Pront) and manually hold in Relay LC 459 CX and manually hold in Relay LC 460 CX. Manual control of Pressurizer heaters and Normal Letdown is now l J available. j k w^. 5.{-[ Operate heaters as necessary to return pressure to normal This is the preferred method of re-energizing the pressurizer heaters per Off-Normal y s, Operating Procedures ONOP.-003.6 and ONOP-003.9. The method listed in the answer key is still approprjate but is now the backup cnethod. ?

REFERENCE:

ONOP 003.6 Page 11 ONOP 003.9 Page 8

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02/06/86 4 P g2 9 NRC Exam Section 4 QUESTION RO 4.19: (RO Regual 4.12, SRO f.21, SRO Requal T.12) A 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.tre the RCPs tripped? RESPONSE:' We request that you modify your answer to state the following: .k 1 1 "The reason for purposely tripping the RCPs during accident conditions is to prevent l 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 j the accident." The basis for our request is that this will reflect the actual wording of the reference stated in the answer key. j .f tr

REFERENCE:

Turkiy' Point Emergency Operating Procedures Student Information Book Section 7, g, page8 1 i k,- ~

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02/06/86 Pcg210 NRC Exam Section 4 k y QUESTION RO 4.20: (SRO 7.22, SRO Requal 7.13) a) What Indication is used in the procedures to denote sub-tasks whleh must be performed in sequence? RESPON8E:- v g, We request that you r.ecept as an additional correct answer to this question to be g; " numbers". The basis for this is found in the document referenced it thr. NRC answer key,' Westinghouse User's Guide, page 3-3 which says; j If sequence of performance is important, then the sub-tasks are oesignated LT etters (or l numbers if finer detall is provided). .,3

REFERENCE:

2 WestlWm User's Guide section 3, page 3 s y-e l Y_'. i l

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Pega 1 NRC Exam Sesti:n 5 g qualmON SRO 5.13: WRO Requal 5.10) What are the four conditions that Tech 8 pees say must be met to ensure the Nuclur Enthalpy Rise Hot Channel Factor is maintained within limits during periods between in-core surveillances? RESPONSES We are requesting that you accept 12 steps as another :::;,-teMe answer. ( 9 The basis for this is that the Techalcal Speelfloations bases state that maintaining the rods within 12 steps precludes a rod misalignment of greater than 15 inches with ~o consideration of maximum instrument error. l b

REFERENCE:

j l T.S. 83.2, Pg. 3.1-5 ] L g a c s. W i l i f

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s.asas Pcto 2 NRC Exam Sesti:n 5 QUESTION SRO 5.18: A rod drops and sticks at the core mid-position from full power conditions with all rods out. A reactor Trip does not occur. If this condition were to persist for an extended period of time (well beyond T/8 limits), what will be the effect on the Excore Axial l Offset of the Power Range NI for the quadrant in whleh the dropped / stuck rod occurs. Include a discussion of menon effects and the definition of axial offset. REEPONEEs k .. p.. :;t 3 We are requesting that you accept the following equation as ano'ther proper definition *of exeore axial offset (EAO) ar EAO = RautorPower l c

REFERENCE:

~ (NTO Reactor Core Control pg. 8-29) \\. ? 4 I i s

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Page3 NRC Ex m Be: tion 5 4 g> QUESTION BRO 5.22: Identify the segments on the actual T-5 Rankine Cycle on the following page for the processes below. (example answers d) segment 5-4) a) Reheat from Moisture 8eparators b) Heat addition from feedwater heaters e) Heat removed by the HP turbine We are regnesting you consider as another correct answer to section B: i h b) 2-3 and the 1st part'of 4-5 e o This answer takes into'secount the high press feedwater heaters, GA and 48. These heaters are on the discharge side of the main feedwater pumps, presented by line 3-4 on g, the graph. REFELMNCE: ~ f 8D-112, Figare 1 d ( d i

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02/06/86 y Page1 NRC Exam Sesti:n 8 h;b QUESTION SRO S.11: WRO Regnal 4.06) j Flu in the blanks in the statement below regarding the spent fuel pool holst k The hoist and bridge controls are interlocked to prevent raising or lowering the load while the is moving. If the upper limit poeltion switch falls to stop a load lift, the will stop the noist a few inches higher. The up' er p limit position swlteh is set so that the bottom noaale of the fuel assembly will i I ensar NER: II We are requesting you dele *e part 2 of this question. This portion of the question is in h error because ED-441s la error. l The upper position Dever) limit switch is designed for emergener out-off only. ~ lt is not j { latentionally used as the normal rdopping devlee. The geared upper limit switch will be h used in normal operation to limit the maximum height to whloh the hoist can be operated. L i s. f This is documented in OP-18304.13 spent fuel pit bridge erane-periodle test. On part 3 of this question we are requesting that you accept the.upender as an .bitional answer because it is the highest obstruction in the spent fuel pit and canal. 1 i i r l 1 y

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M \\ Page 1 NRC Exam Sestlin 7 g; QUESTION BRO f.14 (BRO Requal 7.99) Haw is the RC8 aooled during rzfueling operations with the refueling cavity full if BOTH RHR pumps fall to operatet mESPONSEs We request that you accept as an additional answer that by maintaint.g 23 ft above the I roaster veaul change an adequate heat, sink is provided enough time.to lattiste 3 emergency procedures to ecol the core. f, Y >1 ? mRraRRucas ?, I' s, B.3.19.7 ?. f, v 4 j, s. ) j s. i

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02/06/86 Page 2 NRC Exam Section 7 1@- QUElmON SRO T.15: Besides required notifications, what are the immediate operator actions if you are on l ahlft in the control room and the refueling supervisor in the containment reports they dropped a spent fuel element in containment 7

RESPONSE

I The off-normal procedure for accidents involving new or spent fuel (ONOP-033.3) has been re-lasued. Immediate actions for a dropped spent fuel element are now: I 1. Sound the containment evacuation alarm. 2. Stop the containment purge supply fan. g i 3. Stop the containment purge exhaust fan. J 4. Close the containment purge supply isoL valves POV-3-2600 and 2601. 5. Close the containment purge exhaust isol. valves POV-3-2602 and 2603. 6. Close the containment inst. air bleed valves CV-3-2826 and CV-3-2819. j Subsequent actions include i -1. Place the control room HVAC in the reelreulation mode in accordance with '~ ~ ONOP 10308.1, Control Building Heating Ventilation and Air Conditioning System (HVAC). ' 2. Concurrently perform 3-ONOP-067, Inadvertent Release of Radioactive Gas. y 3 Inform the Plant Supervisor ~ Nuclear to refer to Emergency Procedure 20101, Duties of Emergency Coordinator, and take any actions that may be required. t 4. Notify the Health Physics Supervisor to monitor radiation levels and to evaluate airborne particulate and gaseous samples. 5. Monitor the Containment Atmosphere Process Radiation Monitors R-3-11 and R-3-12 for indication of increasing particulate or gaseous activity. 6. Ver!!y that the containment has been evacuated. We therefore request that the required answers be limited to those immediate actions in the ONOP now in effect.

REFERENCE:

PTN ONOP-033.3, page 6 "R1-5/dsp

02/06/86 4 l Pcg2 3 J l NRC Exam Section 7 QUESTION BRO 7.18: (BRO Requal T.18) Unit 3 is shutdown,4KV Bus 3A is deenergized, a EDG and #3 startup transformer are l teth INOPERABLE. It is required that certain vital loads on Bus 3A be operated. List I three methods (including power source and any interim busses) by which this bus can be reenergized. r, kESPONSE: l We request that you add three additional methods by which the 3A bus may be energized 4 to the answer key. .1 ] 1) B-EDG via the 3B bus via A to B bus tie breaker 2) The switchyard via the main and aux. transformers. (Disconnect links can be ] removed to isolate the main generator) This method has been utilized during outages. { 3 3) The switchyard via the units 1&2 startup transformer via the cranking diesel bus 3 via the 2C 4KV bus to the 3A 4KV bus. l

REFERENCE:

~ C Drawing 5610-T-E-1591 j i y r.

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q2/06/86 Page 1 NRC Ex::m SectlI;n 8 QUESTION SRO 8.04: (8RO Requal 8.03) V-Using the attached technical speelfication, which action below 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 i AFW equipment is OPERABLE) a. Unit 4 must be cooled down to < 350 degrees within 12 hours. b. RITHER Unit 3 OR Unit 4 must be shutdown / cooled down tc, 350 degrees within 72 hours. o e. BOTH Unit 3 and Unit 4 must be shutdown and cooled down to < 350 degrees j within 12 hours if C AFW pump cannot be restored within 72 houra. tL Unit 3 must be shutdown and eccled down to < 350 degrees within it hours if C AFW pump eennot be restored within 72 hours. e. No setion is required. f i

RESPONSE

We feel that answers A or D should be accepted as correct. The reason for our request is stated below. F The foDowing discrepaneles are noted in this question: i 1. Not all the applicable Technical 8peelfications necessary to evaluate this situation were included. Speelfically section 3.0 was not included.- [ 0 s. 2. Unit 4 cannot be in the Startup Mode (Mode 2) when Tave = 410 F. The min. ~ temp. for critically of. Unit is 422'F, per OP-0202.2 pg. 5, paragraph 5.1. i These discrepaneles led candidates to a variety of conclusions. y

Answer A would be the most correct under the following conditions:

4 1. Unit 4 is in Mode 2, Startup mode 2. Unit 3 is in Mode 1, Power operations 0

3.

T.S.3.8.4 (b) is applicable because both units are > 350 F i 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 0 i and to be in Mode 4 within the following 6 hrs thus requiring Unit 4 to be < 350 F 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. Thus one unit must be shutdown and cooled down to < 350 F within 12 hrs if the Inoperable AFW pump cannot be restored within 72 hrs.

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O2/08/88 f Page2 -i NRC Exam Becti:n 8 QU88'110N BRO 8.13: (8RO Requal 8.89) \\ List the three meteorological conditions which would preclude conducting a routine Gaseous Waste release. R88PON8E 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 Northeast. See the attached site plan. h REFERENC8: F8AR Turkey Point Plant Units 3 & 4 Vol.1 Fig. 2.2-3 O l s. .I i

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92/06/86 Pcge 3 f NRC Exam Section 8 QUE8110N SEO S.17: List the four methods available for determining the release rate for Off-8ite Dose calculations and Indicate whleh one is the preferred method. BERPONSE: We request that this question be deleted. The basis for our request is that the Rmergency Coordinator is responsible for directing this activity but the formal responsibility for performing the off-stte does calculations is the responsibility of the ~ redlochemistry department.

REFERENCE:

RPSeite Pg3 Status of Corrective Actions Report PTP-EP-84-OO2 (Response to IE Inspection Report 84-01) I f' O 9 I

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Q2/06/84 f Page 4 NRC Exam Sesti:n 8 QUESTION SRO 8.18: (BRO Requal 8.12) List the DNB related parameters as stated in Tech Spees, and their setpoints. (Assume normal power operations) RERPONSE: The NRC answer key has an apparent typographical error indloating the DNB limit on Pressuriser Pressure in units of psig instead of pala.

REFERENCE:

T.S.3.1.8.b i O L .e 2 I: t i s .l

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02/06/86 F:g3 5 NRC Exam Section 8 ('f s ' QUESUON BRO 8.20: (BRO Requal 8.13) n. a) Where is the access code for the Autodlater retained? b) How is operability of the Autodialer verified prior to being utilized in an Emergency situation? MREPONSE: g 1 We request this question be deleted because no training was provided on the use of the t Autodialer. Training was not conducted on this because the Autodialer has been and j remains inoperable. s I

REFERENCE:

FPL Letter Dated February 24,1983 I C45*: l a +> s. t g - y 1 j

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02/04/88 e Page6 NRC Exam Section 8 (,' QUESTION BRO 8.23: o While performing the Power Range Nuclear Instrumentation Channel Functional Test, a caution states that before taking a UNIT 4 power range channel out of service, two conditions should be verlfled. List these two conditions. i RERFONSE: We submit that an additional correct answer to this question is that it must be vertfled that no other power range channel be la TEST. The Msis for,this is that OP-13304.1, the procedure referenced by the NRC, requires in section 8.1 that the power range channels be tested one at a time. Binee this is a condition that should be verifled prior to taking power range channel out of servlee we feel it should be accepted. R,mR=Nc : 1 OP 12364.3 Sect 8.2 Y i a' s, 4 I i l t i lk

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