ML20235X590

Forwards Comments on Written Senior Reactor Operator/Reactor Operator Exams Administered on 870817

ML20235X590
Person / Time
Site:	Davis Besse
Issue date:	08/26/1987
From:	Shelton D TOLEDO EDISON CO.
To:	Burdick T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
Shared Package
ML20235X498	List: ML20235X497 ML20235X553 ML20235X590
References
1-748, NUDOCS 8710200052
Download: ML20235X590 (179)

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{{#Wiki_filter:- i l TOLEDO EDISON i E DONALD C. SHELTON ~ -/ Vcs presort-4rter (4191 R412399 + Docket No. 50 346 7 ,, g 3 1 Licenr e Mo.. NPF-3 N E2 N Serial ~No. 1-748 !$h F. $:-Y m Y h ' August 26, 1987 h !5', E3 pA o s lt. /, Mr. Thomas Burdick United States Nuclear Regulatory Commission Region III 799 Roosevelt Road, Building 4 Glen Ellyn, IL 60137

Dear Mr. Burdick:

I Enclosed are comments on the written Senior Reactor Operator / Reactor Operator Examinations given at the Davis-Besse Nuclear Power Station, i Unit No. 1 on August 17, 1987. These comments will also be provided to Mr. Tim Reidinger on August 27, 1987. If you have any questions, please contact Mr. Richard Simpkins of my staff. Sincerel yours, / ~(m CAB: pig Enclosure cc: DB-1 NRC Resident Inspector G. Salyers, Region II (SRO comments only) M. Spencer, Idaho National Labs (R0 comments only) B710200052 87tOpg PDR ADDCK 05000346 PDR THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MAOISON AVENUE TOLEDO, OHIO 43652 --__w

c-t i? DAVIS-BESSE R0 EXAM RESPONSE QUESTION 1.03 (2.00) List the FOUR heat transfer regions in an OSTG and indicate how each region changes -(INCREASE, DECREASE OR REMAINS THE SAME) as power is increased from 20 to'100%. ANSVER 1.03 (2.00) 1. Feedvater heating / Preheating INCREASE 2. Nucleate boiling INCREASE 3. Film boiling REMAINS THE SAME 4. Superheat DECREASE REFERENCE 1.03 DAVIS-BESSE Exam Question Bank #01-29 DAVIS-BESSE Training Information Manual, Vol 1, OSTG, pages 18-21 RESPONSE 1.03 Request accepting "Downcomer Aspirating Region" as an alternative for full credit to Feedvater Preheating Region. Aspirating steam is drawn into the Downcomer Region by the feedvater flowing through the feedvater nozzles. Recommend accepting for full credit either answer that the Film Boiling Region remains the same or increases. From 25-100% power, the Film Boiling Region increases two (2) feet, which is a slight increase. When looking at the overall tube length, the change in this region can be considered as remaining the same. REFERENCE PROVIDED Training Information Manual for Once Through Steam Generator i t 1 1 l l Page 1 of 33 l l i _. j

QUESTION 1.04 (1.00) What are FOUR engineering practices used to minimize waterhammer by the operator? ANSVER 1.04 (1.00) 1. Slov opening of valves betvcen voided system and full system. I 2. Proper venting of components prior to starting. 3. Ensuring adequate levels in tanks, where these tanks provide a supply and/or a surge function. 4. Proper use of steam traps and vents. 5. Following operating procedtres. (Any 4 0 0.25 ea. - 1.0) I i REFERENCE 1.04 DAVIS-BESSE Exam Question Bank #01-55 DAVIS-BESSE, HTR-OLC-022-00, Objective 4 061000K505 056000K503 039000K501 193006K104 ..(KA's) RESPONSE 1.04 Recommend the following additional responses be considered for credit: 1. Varming up steam lines before unisolating them. l 2. Starting pumps with discharge valves closed or throttled (following pump starting procedures). REFERENCE PROVIDED j SP 1106.25 l i i l i l l I Page 2 of 33

n l' i OUESTION 1.05 (1.50) Explain HOV and TWO reasons VHY natural circulation flow would be affected by the operator manually raising the OTSG 1evel above the lov i level. limits. ASSUME natural. circulation flow was stable before changing the level. ANSVER 1.05 (1.50) Raising the level vould tend to increase natural circulation (0,5) by raising the effective OTSG thermal center (0.5), and thus increasing the height difference between the heat source and heat sink (0.5). REFERENCE 1.05 DAVIS-BESSE Exam Question Bank #01-34 DAVIS-BESSE, HTR-OLC-011-00, Objectives 1 & 3 193008K123 ..(KA's) RESPONSE 1.05 Recommend you accept for credit respenses that describe the larger density differences that vill occur du* to increased heat' removal in i cold leg as OTSG 1evel is raised, increasing the rate of primary to secon6ary heat transfer. REFE} ENCE PROVIDED j Lesson Plan HT/FF-LP-23 ) i Page 3 of 33

OUESTION 1.06 (2.00) Indicate H0V each of the following items are affected during a RCS heatup. Answer with INCREASE, DECREASE or NO CHANGE. a. RCP head b. RCP power c. RCS volumetric flovrate d. RCS mass flovrate ANSVER 1.06 (2.00) a. NO CHANGE b. DECREASE c. INCREASE d. DECREASE REFERENCE 1.06 Generic: Centrifugal Pumps and System Hydraulics, Igor Karrassik. DAVIS-BESSE, HTR-OLC-019-00, pages 1-15 191004K105 191004K107 193006K101 ..(KA's) RESPONSE 1.06 Recommend accepting the following per the Reactor Coolant Pump Curves. If accept "no change" in RCP head, recommend accepting "no change" in RCP Volumetric Flovrate. If accept an " increase" in RCP Volumetric l Flovrate, recommend accepting a " decrease" in RCP Head as additional responses for full credit. l REFERENCE PROVIDED 1 PP 1101.01, cc 14.70a, Reactor Coolant Pump 1-1

QUESTION 1.08 (1.00) Why should the APSR position remain unchanged, following a reactor i trip? ANSVER-1.08 (1.00) They may. insert positive reactivity _ depending on their position. i REFERENCE 1.08 l DAVIS-BESSE Exam Question Bank #01-106 192005K106 ..(KA's) RESPONSE 1.08 Vording of question may not produce intended answer. This question may produce the physical reasons APSR position remains unchanged. Recommend you accept for credit the six (6) differences listed in l references. REFERENCE PROVIDED Lesson Plan PVR-OLC-036, Exam Questions and a'TP l i Page 5 of 33 _ a _____ ___

l QUESTION 1.09 (2.50) List FIVE of SIX factors which are considered in shutdown margin when the reactor is in a shutdown mode. ANSVER 1.09 (2.50) 1. RCS temperature 2. Boron concentration 3. Xenon concentration 4. Samarium concentration 5. Control rod position 6. Fuel burnup based on gross thermal energy generation (3 @ 0.5 ea

r. 1.5)

REFERENCE 1.09 I DAVIS-BESSE Exam Question Bank #01-116a DAVIS-BESSE, RTR-OLC-023-00 DAVIS-BESSE, Technical Specifications 4.1.1.11.e, page 3/4 1-2 192002K114 ..(KA's) RESPONSE 1.09 Recommend the following additional responses be considered for credit: 1. Whether or not there is a known stuck rod. 2. Samarium and Xenon lumped into one (1) term call transient poisons. REFERENCE PROVIDED Vorksheets 3 and 4 of SP 1103.15 Page 6 of 33

i QUESTION 1.12 (1.00) l I a. What is the approximate 50% BOL equilibrium value for Xenon? b. What is the approximate peak BOL Xenon value for trip from 100%? 'I ANSVER 1.12 (1.00) l a. 2.14% deltak/k +/- 0.1% (0.5) l b. 4.4% deltak/k +/- 0.1% (0.5) REFERENCE 1.12 DAVIS-BESSE, RTR-OLC-021-00, pages 506 Objective RTR-OLC-021-00, #02 192006K102 ..(KA's) RESPONSE 1.12 Question asked for approximate values, yet required an accuracy of 0.1%. Also, value listed does not agree with last Cycle 5 numbers. To address both these concerns, recommend tolerance be expanded to-1 04% delta k/k. REFERENCE PROVIDED Cycle 5 physics data i l l l l l l 1 I a l Page 7 of 33 l l l i 3

] OVESTION 1.14 (1.00) l Exr' 11n HOV and VHY the Source Range indications would behave during j core voiding from a LOCA condition. { .l ANSVER 1.14-(1.00) j i Count rate.voul'd significantly increase (0,5), With (erratic) oscillations due to the loss of neutron moderation I capture before reaching the detectors. (1.0) REFERENCE 1.14 DAVIS-BESSE Exam Question Bank #01-72 191002K117 ..(KA's) RESPONSE 1.14 Recommend grading not require mentioning of capture in response since this is a minor effect when compared to changes in neutron moderation. l l l I 1 1 l l Page 8 of 33

l' I i QUESTION 1.15' (1.50) l u Briefly explain H0V the power coefficient changes over core life, (HORE 1 NEG., LESS NEG., or NO CHANGE) and VHY. ANSVER 1.15 (1.50) HORE NEGATIVE (0.5) -- due to the more negative (0.5) component of moderator temperature coefficient (0.5). f REFERENCE 1.15 DAVIS-BESSE Exam Question Bank #01-09 j DAVIS-BESSE, RTR-OLC-015-00, Obj ective 2 l 192004K108 192004K113 ..(KA's) l RESPONSE 1.15 Question asked for power coefficient changes over core life. Power coefficient does become more negative over core life if you consider the portion of'the curve that is affected by the moderator temperature coefficient (0-28%). The range where Tave'is constant (28-100%) almost no change in the power coefficient occurs over core life. Recommend either answer be acceptable for Iull credit based on assumptions made. REFERENCE PROVIDED PP 1101.02 R0 Curve Book, Figure 14 Page 9 of 33

QUESTION 2.05 (1.50) List the THREE electrical power supplies to the Electrohydraulic ~ Control System. ANSVER 2.05 (1.50) 1. Station Housepover ( vill accept 115 VAC 6011Z ] 2. Permanent Magnetic Generator (PMG) 3 125 VDC electrical distribution system [ vill accept station battery j (3 @ 0.5 ea = 1.5) REFERENCE 2.05 DAVIS-BESSE, TlH Vol 3, EHC, page 19 RESPONSE 2.05 In addition to 115 VAC 60 HZ, recommend accepting for full credit the following for Station Housepover: 480 VAC (E32A or F32A) for the EHC pumps. Also, recommend accepting 120 VAC (YBU) as a source of Station l llousepower. Recommend accepting DBP in addition to 125 VDC electrical distribution since this is the 125 VDC electrical distribution source. REFERENCE PROVIDED SP 1105.12 EHC System Operating Procedure Electrical Distribution Manual Page 10 of 33

i 'i i OUESTION:2.06' (2.00) The Station Air Compressors have several distinct differences. List i FOUR of the instrument and control differences. .I 7 ANSVER 2.06 (2.00) 1. On the Auto Sentry Panel, the.first five local alarms are the same. Sac.1-2 has alarm #6 for excessive vibration SAC 1-1 does not. 1 2; The Auto Sentry Panel alarm #7 for low coo'.ing vater' flow is for SAC 1-1 only, i .l 3. A start signal on SAC 1-1 closes its 480 VAC breaker. SAC 1-2 480 VAC breaker must be closed for operation. A start signal vill not j 'close the breaker. 4. Pressing the "STOP" button in.the control room vill only stop SAC f 1-1 as long as the button is held. To keep it stopped, t..e compressor switch must be placed in " LOCK 0UT". SAC 1-2 can be stopped by placing the control room switch in either "STOP" or j " LOCK 0UT". ~ (4 @ 0.5 ea = 2.0) REFERENCE 2.06 l DAVIS-BESSE, TIM Vol 8, Plant Air System, page 3 '079000K201 ..(KA's) 1 RESPONSE 2.06 j 1 Recommend accepting for full credit the following as additional l differences between the two Station Air Compressors. 1. The Lov Oil Pressure trip setpoints are different 2. The Lead and standby air compressor load and unload setpoints are different J REFERENCE PROVIDED SP 1104.24 Station and Instrument Air System Operating Procedure ] i I i i l l h 1 ^ i I i l Page 11 of 33 ( L - - _ = _ - _.

o i i e l QUESTION 2,08 -(2.00) List the FOUR major differences in ths design / construction of the control rod assemblies and the axial power shaping rod assemblies. ANSVER 2.08 (2.00) 1. 'The active. length of the silver-indium-cadmium neutron absorber material is only the bottom 36 inches. 2. The female couplings of the CRA and the APSRA have a slight dimensional difference. (This.irsures each type of rod can only be coupled to the correct type of drive mechanism.) 3 The section of tubing above the neutron absorber is vented, therefore filled with borated water, reducing differential pressure across the tube vall. 4. The axial power shaping rod drive does not allow the APSRA to drop following a reactor trip. (This prevents possible positive reactivity addition caused by the APSRA axial location.) (4 @ 0.5 ea = 2.0) REFERENCE 2.08 DAVIS-BESSE, PWR-OLC-036.01, Control Rod Drive, page 7 RESPONSE 2,08 Recommend accepting for full credit the following additional differences between the control rod assemblies and the axial power shaping rod assemblies: 1. APGRs have a frictioti screw (or friction button) 2. APSRs have no snubber assembly (snubber includes Bellvile vashers/ springs) 3. APSRs have no flow holes, thus they have no Ball Check Valves 4. APSR bearing surface is thicker (segments) i REFERENCE PROVIDED ) l PVR-OLC-036.01 Control Rod Drive l l i l I l l ) i Page 12 of 33 J

I l [.. QUESTION 2.09 '(1.00) k \\ l-List:FOUR sources available to-the Borated Vater Storage Tank, ANSVER 2.09 (l'. 00) Jl 1. Containmen't Spray pump 1-1 l 2. ' Containment Spray pump 1-2 3. Borated. Water Storage 4. Spent Fuel Pool 5. Fecire, return from HPI pump 1-1 3 6. Recire. return from HPI pump 1-2 l (any 4 @ 0.25 ea - 1.0) REFERENCE 2.09 l DAVIS-BESSE, P & ID M-033, Decay Heat Removal Systen, and Emergency Core Cooling Systems 006000K601 ..(KA's) RESPONSE 2.09 Recommend accepting for full credit the following as additional sources 1 of water to the Borated Vater Storage Tank (BVST). 1. Clean Vaste Receiver Tanks (CVRT) 2, Primary Vater or Primary Vater Storage Tank (PVST) 3. Demin Vater or Domin Vater Storage Tank (DVST) 4. Boric Acid Addition Tanks (BAAT) 5. Concentrate Storage Tank (CST) 6. Refueling Canal 7. Transfer Pit 8. Reactor Coolant System REFERENCE PROVIDED SP 1104.66 Borated Vater Storage Tank Operating Procedure SP 1104.80 Clean Vaste Receiver Tank Operations M-033 Decay Heat Removal System and Emergency Core Cooling Systems M-035 Spent Fuel Pool Cooling System H-045 Chemical Addition System Page 13 of 33

L QUESTION 2.10 (1.00). Vhy is' flow through orifices R06A and R06B of the High Pressure Injection System not required or desired during the " piggyback" mode'? (Two reasons required for f.ull credit). ANSWER 2.10 (1.00) l Cooling is provided by the amount of flow out the. pipe break in the RCS and the recirculation valves close during piggyback. Prevents contaminating BVST from Cont. Sump (any two ) (1.0) REFERENCE 2.10 l DAVIS-BESSE, TIM Vol 6, High Pressure Injection System, page 13. SP 1104.07.19, page 4. (LER 85-013) 006000K605 ..(KA's) RESPONSE 2.10 l Recommend accepting for full credit the following as additional reasons why the recirc valves are closed during piggyback operations (no flov through R06A and 6B). l 1. Prevent contamination of the BVST 2. Prevent loss of water from CTMT building (CTMT Sump) REFERENCE PROVIDED 1 SP 1104.07 High pressure Injection System Procedure - f i f ( i I Page 14 of 33 i

QUESTION.2.14 (2.00) y. A. List the TVO purposes of the Emergency Ventilation System. (Setpoints or-values not required; areas or rooms ARE required) B. List TWO systems from which the Emergency Ventilation System has i the capabi'lity of taking a suction. l ANSVER 2.14 (2.00) A 1. Provide negative pressure (1/4 to 1-1/2 vater) following a LOCA to: a. Annulus b. Mechanical penetration rooms 1-4 c. ECCS rooms i d. Makeup pump. room e. Decay heat cooler room (any three required for full credit) A 2. Reduce airborne. fission product leakage to environment by a filtration process. B 1. Containment Purge System B 2. Radvaste Ventilation System B 3. Fuel Handling Ventilation System (any 2 @ 0.5 ea - 1.0) i l REFERENCE 2.14 { i DAVIS-BESSE, PVR-OLC-048 01, Containment Ventilation, pages 17 and 18. 029000K103 ..(KA's) RESPONSE 2.14 Recommend accepting for full credit an additional purpose of the Emergency Ventilation System providing backup ventilation to the Fuel Handling Ventilation. Per Tech Specs, two Emergency Ventilation Systems servicing the Spent Fuel Pool must be CPERABLE. REFERENCE PROVIDED Tech Spec 3.9.12 Page 15 of 33 l

QUSS7 ION 2.16 (1.75) List the normal power supply by " bus" identification for each of the below: 1. RCP 1-1 2. RCP 2-2 3. Station air compressor 1-1 4.' Emergency air compressor 5. HPI pump 1-2 6. Decay heat Pump B 7. CCV pump 1-1 ANSVER 2.16' (1,75) 1. Bus A 2. Dus B 3. E-3 4. F-13 i 5. D-1 6. C.A.F. 7. C-1 REFERENCE 2.16 DAVIS-BESSE, TIM Vol 1, Reactor Coolant Pumps TIM Vol 6, High Pressure Injection System TIM Vol 6, Decay Heat Removal System TIM Vol 7, Makeup and Purification System TIM Vol 8, Plant Air System TIM Vol 9, Component Cooling Water System 062000K210 ..(KA's) RESPONSE 2.16 The power supply for RCP 2-2 is A Bus, not B Bus. Recommend accepting for full credit D1 for DH Pump B cr C1 if stated as an assumption that DH Pump B is DH Pump 1. We call our pumps DH Pump 1 or DH Pump 2 j REFERENCE PROVIDED Electrical Distribution Manual i ) Page 16 of 33 i

i OUESTION 2.17 (2.00) A. Manual Deluge System is one of the six different types of fire suppression systems used at Davis-Besse. List the remaining five fire suppression systems. (1.25) l B. List the three sources of water for the fire protection systems. (0.75) ANSVER 2.17 (2.00) l i A. 1. Vet sprinkler 2. Pre-action sprinkler 3. Deluge 4. Deluge-vaicr curtain 5. Subsurface foam injection (4 @ 0.25 ea = 1.25) B. 1. Fire water storage tank [ electric fire pump) 2. Intake structure [ diesel fire pump] 3. Fire Department pump connection (3 @ 0.25 ea = 0.75) REFERENCE 2.17 DAVIS-BESSE, PWR-OLC-003.01, Fire Protection / Detection, pages 4, 15-24 086000K302 ..(KA's) RESPONSE 2.17 Recommend accepting for full credit similar wording of ansvers given. l Page 17 of 33 l L____---_____--- _J

2;, y 3 g,., '0VESTION,3.01 -(2.00) t LThe' Integrated Contro1' System-shifts to track mode when any of its stations / control' panels areLplac'ed in manual. List.FOUR=other- - conditions or events causing the ICS.to shift to' track' ~ . ANSWER'3.01 .(2.00) l

1.

ITurbine EHC' Control.N0T in ICS auto 2, - Reactoritrip.- 3. BOTH turbine generator output. breakers open (ie. load rejection) 4. Gross limits: 5. Turbine trip (5 @ 0.4 ='2.0) REFERENCE 3.01 ~l . DAVIS-BESSE Examination Question Bank,. Question # 03-6 RESPONSE-3.01 Recommend accepting for full credit in a'ddition to the answers provided, a Reactor Cross Limit and a Feedwater Cross limit if listed separately. The ICS: procedure considers these as separate conditions ovhich vill cause the ICS to-shift to' track. ~ REFERENCE PROVIDED SP 1105.04 ICS Operating Procedure l 1 l l I {.. ( Page 18 of 33 ] A L__ -_2_ __. -_.._

i' I QUESTION 3.02' (2.00) State FIVE of the:seven.RCP starting interlocks. Setpoints are NOT

required.

ANSVER 3.02. -(2.00) 1. Lift oil pressure 2. CCV flov-3. Reservoir oil-level 4. Seal injection flov 5. Reactor power i 6. Core Lift Criteria [ .7. Seal return valve open (5 @ 0.4 = 2.0) REFERENCE 3.02 DAVIS-BESSE, PVR-OLC-033, p. 003000K406 003000K403 003000K402 ..(KA's) RESPONSE 3.02 l Recommend full credit in addition-to the ansvers given for the following: 1. Voltage (>75% per procedure) 2. CCW flow to upper bearing cooler 3. CCW flow to lower bearing cooler 4. CCW flow to motor air cooler 5. CCV flow to RCP seal cooler 1 Per the procedure and the P&ID, the CCV flow to each of the above RCP components is a separate interlock. Also recommend accepting for full credit similar words for core lift criteria such as T' greater than setpoint (500*F). REFERENCE PROVIDED SP 1103.06 P&ID H-040B j i I i l Page 19 of 33 l 1

g -l QUESTION 3.03 (2.00) 1 i State the FOUR ICS runbacks, their rates and their limits. f ANSVER 3.03 (2.00) (0.3) for runbdek (0.1) for rate (0.1) for limit 1. Deaerator Level 50%/ min 55% 2. Loss of RCP 50%/ min 75/45% I 3. Loss of FVP 50%/ min 55% l 4. FVP'Hi Dis. Press b%/ min 55% i REFERENCE 3.03 J

• • • • CAF**** FACILITY DID NOT SUPPLY LESSON PLANS OR TRAINING TEXT!!

DAVIS-BESSE Examination Question Bank, Question #06-43 RESPONSE 3.03 Recoini,:and accepting for full credit that the ICS runback rate for FV High Discharge Pressure is 20% per minute. REFERENCE PROVIDED 1 Training Information Manual, Volume 3, ICS l Page 20 of 33 { 1

l i OUESTION 3.05 (1.50). What POUR trips are active on a. Loss of Voltage or SA start condition on the Emergency l;riesel Generator? 1 f (Setpoints not required) ANSVER 3.05 -(1.50) 1. 0verspeed 2. DG differential relay action 3. Overcurrent 4. Manual Emergency Shutdown (4 @ 0.5 ea = 1.0) .l REFERENCE 3.05 DAVIS-BESSE,.PVR-OLC-055.01, p. 4 Obj ec tives, PVR-OLC-055.01, #04 064000K401 ..(KA's) RESPONSE 3.05 Request that only three.(3) of four (4) ansvers be required for full credit since the procedure only lists three (3) trips that are active on a loss of voltage or SA start condition. Ve don't normally consider the manual emergency shutdown a trip. REFERENCE PROVIDED SP 1107.11 Diesel Generator Operating Procedure I l 1 1 1 i Page 21 of 33 i

_ _ _ _ _ = _ _ _ _ _ _ _ _ _ _ _ _ - _ _ f. ? QUESTION 3.07 (1.00) ^ 't Vhat must be done on the control board to open either Auxiliary Feed Pump Turbine steam supp'.y valve, if lov AFP suction pressure condition exist? ANSVER 3.07 (1.00) I Depress and HOLD the open push-buttons. REFERENCE 3.07 DAVISLBESSE, PVR-OLC-029.02, p. 12 Obj ec tives, PVR-OLC-029.02, #06 061000K414 ..(KA's) RESPONSE 3.07 Recommend accepting for full credit in addition to answer given that the Service Water Isolations to the Aux. Feedvater Pumps (SV 1382/ 1383) should be opened. This vill clear the low pressure trip on the steam supply valves. The CTRM operator can now open these valves by j pushing open the pushbuttons for the respective valves. We vouldn't open the steam supply valves during a lov suction pressure condition because this would result in Aux. Feedvater Pump damage. Page 22 of 33 1

. -y yn' [! QUESTIONL3.08' (1.00) y J' ) 1 i 'What are-fthe automatic actions'that occur;in;the' event the setpoint is reached on'RE-1822 A or B (Vaste. gas)?. ANSVERl3.08. (1.00) -Plant Vent isolation' valves. shut (VG'1819 6 1820).. REFERENCE 3.08 ) I DAVIS-BESSE,'PVR-0LC-004.01, p. 17 Objectives, PVR-OLC-004.01, #05 071000K404- ..(KA's)- It'. -RESPONSE 3.08

-Recommend' full credit in. addition to the answer given for the following

' automatic: actions. 'An alarm on RE 1822A or B will result in an' alarm .on the CTRM RMS/FDS console and the CTRM annunciator " Unit Fire or Radiation' Trouble".1 REFERENCE PROVIDED AP 3009.24 Unit Fire or Radiation Trouble Alarm Procedure. P&ID'M-0381 Gaseous Radwaste System I i I I [ Page 23 of 33

QUESTIONf3.10 ~ (2.00)- NAME the TVO cross limits associated with the ICS and BRIEFLY describe each' cross limit. Include in'your discussion, the conditions under which the limit vill be in effect and the demand signal (s) which1 rill 'be modified by the limit (including direction of change). 1 ANSVER 3.10 - ( 2. 00.) 1. 'FEEDVATER CROSS LIMIT (0.25) If measured feedvater flow deviates from feedvater demand by >/= -5% (0.25), reactor demand vill be modified by 1% in excess of 5% (0.25). Reactor demand will only be decreased by the actions of this limiter (0.25)'. 2. REACTOR CROSS LIMIT (0.25) If measured reactor power deviates from reactor demand by -5% to +10% (deadband) (0.25), then feedvater demand vill be modified by the amount of error in excess of deadband (0.25). Feedvater demand may be increased or decreased to match reactor power (0.25). i -REFERENCE 3.10

• • • • CAF**** FACILITY DID NOT SUPPLY LESSON PLANS OR TRAINING TEXT!!

059000K402 ..(KA's). RESPONSE 3.10 l Additional information has been provided for the Reactor Cross Limit. For every 1% above the 410% deadband value, the FV demand is increased by 1/2%. Fore every 1% below the -5% deadband value, the FV demand is decreased by 1%. REFERENCE PROVIDED Training Information Manual, Volume 3, ICS j l Page 24 of 33 L_-______----.-_________-._____

T 's p V L QUESTION 3.17 (2.50) +. A. What FOUR actions: vill take place if a " Bus Lockout" occurs on the C1 essential bus? (2.0) B. Wht action (s) occur as a result of depressing the " Lockout Reset" button? .(0.5) ANSVER 3.17 (2.50) A. 1. Normally closed supply breaker opens. 2. EDG starts but does not tie to bus. 3. The other TVO supply; breakers get a " prevent close" signal. 4. All load breakers trip. (4 @ 0.5 ea - 2.0) B. The EDG output breaker closes on the bus. (0.5) REFERENCE 3.17 DAVIS BESSE, PVR-0LC-054.01, pages 8 Objectives, PVR-OLC-054.01, #04 062000K403 062000K402 ..(KA'S) RESPONSE 3.17 Part A: The Emergency Diesel Generator must be shutdown since it will ) be running without component cooling vater available.. Recommend that i this response be accepted in addition to the four (4) listed for 3 i credit. 1 REFERENCE PROVIDED j l SP 1107.05 4160 Volt-Svitching Procedure i F 1 i Page 25 of 33 l l 1

L@ . r.y!~. L L >t Jr-d.; i y '[i'00ESTION4.02 (1.00) According to Safety Tagging AD 1803.00, restoration to service shall not begin until what THREE items are completed? ANSVER. 4.02 (1.00)- 1. Vork has been completed 2 '. All personnel holding clearance have released clearance 3. The SS/ Assistant SS have determined the equipment is ready for service. (3 @ 0.33 ea = 1.0) REFERENCE 4.02 -) DAVIS,-BESSE, ADM-SRO-005.01, Safety Tagging Procedure AD 1803.00 194001K102 ..(KA's) l m RESPONSE 4.02' j 1.< Determining required tests or inspections Request additional responses be considered for full credit. 2.- Safety Tags have been removed 1 1 3. 'Valvc lineups required for return to service 4. Electrical lineups required for return to service 5. System filled and vented 6. Maintenance personnel present for post maintenance running of equipment 7. Satisfactory disposition of any PCAO or NCR 8. MVO signed off REFERENCE PROVIDED l AD 1839.00 Station Operations AD 1844.02 Control of Work AD 1803.00 Safety Tagging-i i l 1 l Page 26 of 33

l. I l. QUESTION 4.03 (1.00) i 1. List FOUR of.the.seven Reactor Operator responsibilities as i identifie6 i AD 1839.00, Station Operations. s ANSVER 4.03 (1.00) l l 1. Operating the Nuclear Steam Supply System, turbine generator, l their auxiliaries,.and all of the equipment to 7aintain continuous-I[ production with maximum safety and efficie,ncy.3, t 2. Reporting to the Assist. SS information importa'nt'to the safe and efficient operation of the station. -). / ^ l -3. Being a) art and attentive to the instrumentation and controis in the control room and control room' area, i.>' 4. Maintaining control of personnel entering the control room. 5. Maintaining the Reactor Operator's Log, control room reading sheets, and the Primary Plant Status Board. ) ~ 6. Shutting down the reactor when he determines that the saUety jf i the reactor is in jeopardy or when operating parameters exceed any reactor protection setpoints and automatic shutdown does not occur. j 7. Maintaining as NRC Reactor Operator's license on the unit. (any 4 @ 0.25 ea = 1.0) REFERENCE 4.03 1 DAVIS-BESSE, AD 1829.00, Shift Operations, page 9 002000SG01 ..(KA's) RESPONSE 4.03 AD 1839.00 was recently revised which increased the responsibilities of the Reactor Operators. Recommend accepting for full credit the expanded list of responsibilities for Reactor Operators listed in the applicable portion of AD 1839.00. See attached reference. REFERENCE PROVIDED AD 1839.00 Station Operations Page 27 of 33

l {- l QUESTION 4.06 (1.50) HP 1601.01.10, Guide and Limits for Exposure to Radiation places three limits or guides on quarterly and annual exposures, 1. What is the working quarterly guide limit? 2. What is the working quarterly maximum allowable limit? 3. What is the working annual guide limit? ANSVER 4.06 (1.50) 1. 1250 mrem 2. 3000 mrem 3. 4500 mrem (3 @ 0.5 ea = 1.5) REFERENCE 4.06 DAVIS-BESSE, HP 1601.01.10, page 2 ) 194001K104 ..(KA's) RESPONSE 4.06 l'6 Recommend full credit be given for 5 rem per year as the working annual . ); ,/ guide limit in addition to the answer given. HP 1601.01 states that t the working quarterly guide limit of 1250 mrem per quarter is equivalent to 5 rem per year. s REFERENCE PROVIDED 4 HP 1601.01 Guides and Limits for Exposure to Radiation i l Pnge 28 of 33

l L 00ESTION'4.07 (2.00) SP 1103.04.8, Boron Concentration Control, list several requirements to stop boration or deboration. List FOUR of these requirements. ANSYCA 4.07 (2.00) 1. Control rod safety group 1 is at'its upper limit ~ 2. The makeup and purification system is disrupted by a loss of letdown, loss of makeup pump, loss of makeup flow, etc. 3. If control rod group position indication, neutron count rate, or other reactivity indications are behaving in an erratic unexpected manner. 4. If the' time calculated for the operation varies significantly from actual operation time. (4 @ 0.5 ea = 2.0) REFERENCE 4.07 DAVIS-BESSE, SP 1103.04.8, Boron Concentration Control, page 2 004000SG01 ..(KA's) RESPONSE 4.07 Tech Specs require the suspension of all operations involved in the reduction of boron concentration of the RCS when certain conditions are not met. Some of the Tech Specs are listed in the Precautions an. Limitations of SP 1103.04. Recommend accepting for full credit, it, addition to the conditions given, all Tech Specs which direct actions for stopping decorations. See references. REFERENCE PROVIDED SP 1103.04 Boron Concentration Control T.S. 3.1.1.2 T.S. 3.1.2 T.S. 3.1.2.3 T.S. 3.1.2.6 T.S. 3.1.2.8 i T.S. 3.8.1.2 T.S. 3.9.1 T.S. 3.9.2 Page 29 of 33 ) l .1

QUESTION 4.09 (2.00) List the FOUR actions required of the Reactor Operator per AB 1203.12, Control Room Evacuation, before leaving the Control Room. ANSVER 4.09 (2.00) 1. Start a second makeup pump i 2. Isolate letdown 3. Manually trip reactor 4. Verify all control rods inserted (except AFRS) (4 @ 0.5 ea - 2.0) REFERENCE 4.09 DAVIS-BESSE, AB 1203.12, page 5 [47] 002000SG14 ..(KA's) RESPONSE 4.09 The question doesn't specify whether the actions required are those for the Prir..ary RO or Secondary R0. Recommend accepting for full credit the additional actions taken by the s:condary R0. 1. Trip SFRCS on high level 2. Verify proper SFRCS response 3. Block the SFRCS signal to both Atmospheric Vent Valves and place i them in auto 4. Open condenser vacuum brkrs. 5. Lockout Hechanical Hogger 6. Take emergency key ring l l Page 30 of 33

'(2.00) QUESTION 4.12 1 A. Adequate subcooling margin exists when the t-sat meter indicates greater than or equal to which of the below? A. 10 degrees F B. 15' degrees F i C. 20 degrees F D. 35 degrees F (0.5) B. According to the Emergency Procedure, EP 1202.01, Specific Rule 1, l what THREE actions are required to be performed on the makeup i system in the event of a loss of adequate subcooling margin? i ANSVER 4.12 (2.00) C (0.5) A. = B. 1. start a second makeup pump 2. fully open MU-32 (makeup control valve) 3. shift makeup pump suction to the BVST by closing MU-3971 (makeup pump 3-vay saction. valve) l l (3'@ 0.5 ea = 1.5) REFERENCE 4.12 DAVIS-BESSE, G0P-OLC-003, OBJ 1 AND EP 1202.01 [9] [11} 10CFR5521J ..(KA's) RESPONSE 4.12 Recommend accepting for full credit either closing MU 3971 or shifting MU 3971 to the BVST position. The key lists closing MU 3971 while EP 1202.01 says shift to the BVST position. REFERENCE PROVIDED EP 1202.01 Specific Rule 1 l l 1 i f' Page 31 of 33 l i

L QUESTION 4.13 (1.00) l List the FOUR duties of the Primary and' Secondary Reactor Operators. g. (EP 1202.01) ANSVER 4.13 (1.00) 1. Carries out immediate actions as directed by applicable procedure 2. Performs actions as directed by Assistant Shift Supervisor 3. Acknowledges in a positive manner the completion of actions directed by the " Procedure Reader" during implementation of EP 1202.01 4. Made appropriate recommendations to Assistant or Shift Supervisor. REFERENCE 4.13-DAVIS-BESSE, G0P-0LC-003.00, Page 15 194001K109 ..(KA's) RESPONSE 4.13 During the exam several candidates asked for clarification to this question. Some candidates were told that these duties didn't necessarily apply to EP 1202.01, but were operator duties that would be applicable when entering any procedure, such as an Alarm Procedure (AP) or an Abnormal Procedure (AB). As such, we recommend a vide latitude of duties be acceptable for full credit. Some duties have been . attached for your information. REFERENCE PROVIDED AD 1839.00 Station Operations 1 1 i ) Page 32 nf 33 I 4

1 i l l OUESTION 4.18' (1.00) 1. What, approximate power level on the intermediate range instrument ' corresponds to the point of adding heat? (0.5) 1 2. What value should startup rate (SUR) be below prior to the point i .of adding heat? (0.5) 1 ANSVER 4.18 .(1.00) ) 1. Between 5 x 10(-8) and 5 x 10(-7) amps. j i 2. Less than 0.1 DPH (2 @ 0.5 ea = 1.0) REFERENCE 4.18 DAVIS-BESSE, Plant Startup, PP 1102.02.25, page 66 010000SG07 ..(KA's) RESPONSE 4.18 Recommend full credit for a SUR of approximately 0.1 DPH (per PP 1102.02) in addition to the less than 0.1 DPH given in the answer key. REFERENCE PROVIDED PP.1102.02 Plant Startup Page 33 of 33

' DAVIS-BESSE SRO EXAM RESPONSE QUESTION 5.04 (1.25) -f i

1.

How is Xenon-135 produced and removed in the core? Give decay { times. (1.0) 2.. What percent of all fissions, either directly or through decay, produce Xenon? (0.25) ANSVER 5'.04' (1.25) 1. From decay 52 Te 135 B- (2 mi) 53 I 135 B- (6.6 hr) 4 54 Xe 135 B-(9.2 hr) 55 Cs 135 (0.5) Some from direct fission. Removal is from decay and burnup. (0.5) 2. 6.3% +.3% (0.25) . REFERENCE TECO HTT-SRO-010 pg 12 19200KS112 ...(KA'S) RESPONSE 5.04 Distribution'of points for Part 1 is unclear. Recommend even . distribution of points for each.of the production / removal terms. Part.2 tolerance on answer is very restrictive. Request a band of 5-7%. It is more important that we understand that this represents a small portion of the total Xenon production, rather than exactly 6.3%. l Page 1 of 31 l-x

i l I i QUESTION 5.07 (2.00)- 1.. Define SHU'"9VN MARGIN per T.S. (1.0) i 2. List 4 of the 5 parameters for calculating SHUTD0VN MARGIN as listed in SP-1103.15 worksheet 4. (1.0) i ANSVER 5.07 (2.00) 1.- The amount of reactivity that the reactor is or instantaneously could be made subcritical from its present condition assuming no change in APSR position, and all control rods are fully inserted i except for the single rod of highest worth which is assumed to be fully withdrawn. 2. Red worth, reactivity of temp, boron, xenon, fuel (0.25 ea. total of'1.0) i-REFERENCE TECO RTR-OLC-023 192002K13 ...(KA'S) l RESPONSE 5.07 The Part 2 list of acceptable parameters should also include Samarium, any known stuck rods, and transient poisons (Xenon and Samarium lumped together). REFERENCE PROVIDED Vorksheet 4 of SP 1103.15 Page 2 of 31 m____________. J

L;p-o-)' 3 6 %,4 ,1 s \\ . ( ?0UESTION 5.08' '(2.25) 1.'- Define Beta core. (0.5)- 2.- LIST ctlie:following isotopes U235, U238, and Pu239 in ' descending : order of the value of theirsbeta. (0.75) 4- -3. Explain hov and why beta core changer over core'~1ife. (1.0) ANSVER 5.08: (2.25) 1 ~. Beta core-is the sum of all the Bi values for fission of a-par!.icular. fuel'or the_ total fraction of: delayed. neutrons. produced by f.tssion of a particular fuel. (Bi is the fract' ion ofL otal number of neutrons present in'a t reactor which are produced.by a given delayed neutron precursor.) 2.- B-U238 = 0.0156 - 29=b0 1 i 1 3. Beta core decreases over core life due to: U235 U238 Pu239 Power BOL' 93T +--3% 7% +- 3% .0% m Fraction EOL 55l;:+- 5% 7% +- 3% 38% +- 5% (1.0) REFERENCE TEC0 RTR-OLC-'J10 TP 10.6 & 10.7, pg 4 192003K107 ...(KA'S) a i RESPONSE 5 08 Part 1 definition of Beta Core does not address the fact that Beta Core ~ is a veighted average. Recommend Beta Core definition provided in reference be accepted for full'eredit. REFERENCE PROVIDED ] l RTR-OLC-010, TP 10.6 RESPONSE 5.08 Parts 2 and 3 do not ask for numerical values. Recommend full credit 1 be given without describing numerical values. l f Page 3 of 31 umi. j ^

A f, 7 g.-: .n,

)

(, 00ESTIONL5!.10L

(2.00)'

o -B&W 177-fuel. assembly plants have 52 SPND detector' strings.~.The '3 H>, detector' strings are' inserted in the central instrument tube of the. L' '.sel'ected fuel assemblies. N j 1 =. Listiths threeL(three) major components and their quantity in a 'j single'SPND' detector. string. L(1.5)~ yl ' M' . 2." EXPLAIN hov a'SPND detects'.and indicates'a neutron flux?. ~ (0,5) ANSVER 5'.10. .(2.00) l ~1. 1. 'seven (0.25): rhodium emittersL(0.25) a H 2. one (0.25) background wire (detector) (0.25) 3. -one (0.25) thermocouple (0.25) E2. When a rhodium atom in.the emitter absorbs a neutron, (0.1) it is ' transmitted to. rhodium-104,-(0.1) which B-decays producin'g an' electron which produces a current flow. (0.1) this current flow is i ~ -carried by the lead wire to the Control Room. (0.1) The background vire is similar to.a leadvire and is used to account. for gamma-induced currents. (0.1)- (103Rh +;N.---104Rh----194Pd + B-) REPERENCE TECO PVR-OLC-038 015000K501'...(KA'S) RESPONSE 5.10 Recommend changing the wording of Part:1 for future exams'to " List the -3 (three) major components and-their quantity in a single incore detector string." - Requesting the 3 components in an SPND detector string' vill tend to produce responses that deal only with the rhodium i portion of an incore string. r. 'I i i 7.. l Page 4 of 31 j 1

QUESTION 5.12 (1.50) LIST 3 of the 4 methods itsed to verify that natural circulation heat removal to the OTSGs is present according to PP 1102.10 Plant Shutdown and Cooldown (include limits as appropriate). (1.5) i ANSVER 5.12 (1.5') i 1. RCS delta T has stabilized, does not exceed 50 deg. F. (0.5) 2. Verify heat removal from the OTSGs by observing turbine bypass -valve position or atmospheric vent valve position, (0.25) and auxiliary feedvater flow. (0.25) i 3. Incore. thermocouple temperatures stabilize. j (0.5) 4. The RCS is at least 50 deg. F subcooled per Tsat meters. (0.5) (ANY 3 @ 0.5 ea.) REFERENCE 193008K122 ...(KA'S) RESPONSE 5.12 Recommend accepting for answer 4, P/T display or manual plot as acceptable. REFERENCE PROVIDED EP 1202.01 l I Page 5 of 31 I (.

i i l QUESTION 5.16 (1.50) Given the following data: Th 606 deg. P ~ core output 2700'MV Tc 554 deg. F .Tave 580 deg. F-superheat.35 deg. F 0TSG Header pressure 900# Temperature of the feedvater 580 deg. F OTSG Levels 88% 1. Calculate the primary coolant flow. ANSVER 5.16 (1.50) Opri 1. Mpri - --------- (0.75 for equation) (hh - he) 2700 MV (3.41 btu /v) Mpri = -------------------- -(0.5 for correct "h" determination) 622.8 - 552.3 btu /lbm 9207 E6 btu Mpri = ------------ (0.25 for calculation) 70.5 btu /lbm Mpri = 130.6 E6 lbm/hr + 10E6 4 (1.5) l REFERENCE { TECO HTR-OLC-023 193007K108 ...(KA'S) RESPONSE 5.16 j Recommend for full credit using 0 - M Cp (delta T) as a method of 1 calculating m if cp assumed = 1.3 then m = 136.2 mPPH. Also recommend a band of 10% for all calculations, since no tolerances are sta'ted in the answer. l REFERENCES PROVIDED Typical Cp values from HTFF TP 19.2; steam tables could also be used. l l Page 6 of 31 l L l

QUESTION 5.17 (2.25) 1. Explain what is occurring in the following heat transfer regions 1. ' Nucleate boil'ing region (1.5) 2.- ' Film boiling region 3. Superheat region 2. -As'pover is increased from 0% to 100% power, state whether the area.of the heat transfer. region (increases, decreases, or stays the same) 1. Nucleate boiling region (0.75) .2. Film boiling region 3. Superheat region l ANSVER 5.17 (2.25) 1. 1. Formation of bubbles at imperfections on the surface of the tube. (0.2) Heat is concentrated at the imperfection. Bubbles'are swept away from the tube into coolant. (0.1) The . bubbles agitate the coolant, then collapse. (0.2) (As boiling rate increases you reach bulk boiling. Here the steam bubbles do not collapse.) 2. .The entire surface is covered by a film of vapor, (0.3) which is a poor conductor of heat and acts as more of an insulator than the liquid it replaced. (0.2) (Energy in this region is by radiation and convection.) 3. Enough heat is added to completely vaporize all of the liquid present. (0.3) The further addition of heet causes the vapor to become superheated. (0.2) 2. 1. increases l 2. stays the same 3. decreases REFERENCE j TECO PVR-OLC-026 & HTR-OLC-005 193008K103 ...(KA'S) RESPONSE 5.17 ] Part 2 for film boiling region recommend accepting for full credit increases slightly or stays the same. ] REFERENCES PROVIDED l ) L OTSG Training Information Manual 1 l I l I Page 7 of 31 i i l 1

l QUESTION 5.18 (1.0) the'four (4) th'rmodynamic conditions needed for natural I List e circulation to occur. ANSVER 5.18 -(1.0)

1.

Must have a heat sink. (0.25) 2. Must have a heat source. (0.25) 3. An elevation difference between the heat source and the heat sink. (0.25) 4. Coupling of a subcooled liquid between the heat sink and the heat source. (0.25) REFERENCE TECO HTR-OLC-11 193008K121 ...(KA'S) { i RESPONSE 15.18 Recommend accepting for full credit to Answer 4, a flovpath between heat source and heat sink. Also for Answer 4, liquid does not have to be subcooled. See referenced material. REFERENCE PROVIDED Technical Bases Document for EP 1202.01 l l 1 i l I Page 8 of 31 i l

? 00ESTION 6.01 -(1.00), i Hatch the following SFRCS equipment to their power supply. a.~ SFRCS~ CHANNEL 1 a. YAU (1.0) b. Y1 b. SFRCS CHANNEL 4 c. D1N. e. D1P c. SGLIC 2 f. YBU g. Y2 d. SGLIC 3 h. D2N i 1. D2P (4 0 0.25) j ANSVER 6.01 (1.00) a. SFRCS CHANNEL 1------b (1.0) d. SFRCS CHANNEL 4------i i f. SGLIC 2--------------g g. SGLIC 3--------------e ~ REFERENCE DAVIS BESSE TRAINING INFORMATION MANUAL VOLUME 5, SFRCS 061000K201 ...(KA'S) RESPONSE 6.01 Lettering to the left of responses in answer key doesn't match exam, i but ansvers are correct. i 1 i l l I l 1 i Page 9 of 31

OUESTION 6.02 (1.25) LIST the (five) 5 parameters that vill cause a SFRCS FULL TRIP, including setpoints if applicable. ANSVER 6.02 (1.25) 1. Lov Main Steam Pressure -612 psig i 2. High SC/Feedvater D/P -177 psid i 3. High SG Level -280" 4.= Lov SG Level - 26.5" 5. Loss of four (4) RCPs. (0.15 for each item; 0.10 for each setpoint) REFERENCE DAVIS DESSE TRAINING INFORMATION MANUAL VOLUME 5, SFRCS 061000K402 ...(KA'S) RESPONSE 6.02 Recommend accepting for full credit to Answer #3, 94% on operate range in addition to SFRCS indication. Operate range is available to operator in Control Room. REFERENCE PROVIDED EP 1202.01 i i 1 l Page 10 of 31 L:

l OUESTION 6.04 (1.00) There are (tvo)'2 identical CTHT Hydrogen Analyzers', each capable of sampling one of four different points. STATE vhere these sample points are physically located in containment. ANSVER 6.04 (1.00) 1 (656' El.) - Top of #1 Steam Generator secondary. shield vall (Normal for 1-1) 2. (603' El.) - Personnel Air Lock area (Normal for 1-2) I 3. (817' El.) - Top of_the CTMT Dome 4. (656; El.) - Top of #2 Steam Generator secondary shield vall (4 @ 0.25 ea.)

l DAVIS BESSE TRAINING INFORMATION MANUAL VOLUME 7, HYDROGEN REMOVAL SYSTEM CTMT 028000A403

...~ (l'.A ' S ) RESPONSE 6.04 Steam Generator secondary shield wall commonly referred to as D-rings. Recommend accepting top of each side D-rings for full credit in addition to answer key response. REFERENCE PROVIDED PP 1102.02 Plant Startup Page 11 of 31 )

QUESTION 6.06 (2.00) 1. LIST the.three (3) inputs to ARTS and from where the signal is being sensed, other than SFRCS. (1.5) q 2. VHICH one of the SFRCS actuation signals will not cause an ARTS l trip? (0.5) ANSVER 6.06 (2.00) 1 -1. 1. Turbine trip signal -Emergency trip system oil pressure. (on. the main stop valves less than 275 psig in the turbine EHC). i (0.5) 2. Main feedvater pump trip signal = control oil pressure (less than 75 psig on both pumps) (0.5) 3. Out of core flux signal - from RPS (0.5) 2. Loss of all RCPs (0,5) REFERENCE TECO PUR-OLC-040 SP 1105.21 DAVIS BESSE TRAINING INFORMATION MANUAL VOLUME 6, ARTS 012000K402 ...(KA's) RESPONSE 6.06 Question does not ask for setpoints, while answer key providea setpoints. Recommend not requiring setpoints for full credit. Part 2 answer is now incorrect due to modifications made to SFRCS and ARTS Systems. All SFRCS trips now trip ARTS. Recommend accepting this in addition to key answer for full credit. REFERENCE PROVIDED SP 1105.16 Page 12 of 31

) -0UESTION 6.08 (1.50)- q The HSR' drain tank has a high level turbine trip associated with it. i This trip is 3 inches below the bottom of the HSR. 1. State'the purpose of this trip. (0.5) 2. Describe how the steam / water gets from the HSR entry point'to the HSR drain tank. (0,5) 3. State where the normal level control valve for the HSR drain tank ' drains to. (0.25) 4. State where the high level dump valve for the HSR drain tank dumps. (0.25) ANSVER 6.08 (1.50) 1. Prevents backing water into the HP turbine, which could cause turbine blade damage. (0,5) 2. (Holsture from the steam enters the HSR from the exhaust of the HP-turbine.) As the steam enters it first passes over chevron plates that remove moisture (0.25) (by causing abrupt changes in direction.) This moisture drains down from the plates in the HSR j into the HSR drain tank. (0.25) 3. Feedvater Heater 5-1 or 2 4. Condenser REFERENCE P&ID H005 1 TECO PVR-OLC-028 l 039000K105 ...(KA'S) RESPONSE 6.08 Part 1: recommend accepting for full credit an alternate response of prevents carryover into LP turbine per reference provided. REFERENCE PROVIDED Lesson Plan PVR-OLC-028 i l l r Page 13 of 31 f

QUESTION 6.10-(1.50) LIST five (5) of the eight (8). trips.that will cause an automatic shutdown of the Control Room Ventilation System. (Setpoints not required.) (1.5) ANSVER 6.10 (1.50) 1. SFAS Level 1 Actuation 2. Station Vent High Radiation Level 3. High Chlorine (5 ppm @ ventilation inlet or blockhouse) 4. Loss of control air (<75 psig) 5. Loss of control power 6. Freeze stats'(35 deg. F @ outlet of heating coli) j 7.. Smoke Stats (3% smoke density @ outlet of Air Handling Unit or inlet to.the Return Air. Fan) 8. Fire ~ Stats (135 deg. F @ outlet of Air Handling Unit or 165 deg. F @ inlet to the Return Air Fan) i (5 0 0.3 ea.) l REFERENCE I TECO PWR-OLC-009 DAVIS BESSE TRAINING INFORHTION MANUAL VOL. 8 CONTROL R00H VENTILATION 000060EK30 ...(KA'S) RESPONSE 6.10 .l Request additional responses be considered for full credit: 1. Separate each of the chlorine detectors since they detect at different locations. l 2. Closure of supply and return dampers which causes supply and return fans to trip. REFERENCE PROVIDFD I SP 1104.14, Pages 9 and 1 I i 0 l l-I 1 Page 14 of 31 i i J

QUESTION 6.12 (1.00) LIST.the four (4) trips which are removed when'the RPS is placed into " Shutdown Bypass" mode. (1.0) ANSVER 6.12 -(1.00) ,1. Power / Imbalance'/ Flow Trip 2. Power / Pump Trip 3. Pressure / Temperature Trip 4. Low Pressure Trip REFERENCE TECO PVR-OLC-039 .SP-1105.02 012000K604 ...(KA'S) RESPONSE 6.12 Recommend accepting for full credit alternate names for Trips 1 and 2: 1. Pover/ Imbalance /Plov also known as Flux / delta Flux /Flov 2. Pover/ Pump Trip also known as High Flux / Number of_RC Pumps on REFERENCE PROVIDED SP 1105.02 i l I I I J l ft i l' l Page 15 of 31

s< QUESTION 6.14 (1.00) 1. What' one' system is the major entry point for oxygen being introduced into the Gaseous Radioactive Vaste System? (0.5) 2. Fill in the blank: -If radiation-monitor sample flov drops belov 6. + 1.2 SLPM, an alarm will actuate except when HV 1819 and HV 1820 are 1 (0.5) ANSVER 6.14 (1.00) 1. Clean Vaste System 2. Closed 1 REFERENCE TECO PVR-0LC-004 SP 1104.27 071000A429 ...(KA'S) RESPONSE 6.14 Clean Vaste System is comprised of many components, all of which should be acceptable as a response. Also, for Part 2, our normal practice is not to memorize valve numbers; therefore, valve names should be provided in addition to valve numbers. REFERENCE PROVIDED SP 1104.29 l Page 16 of 31

l f OUESTION 6.15' '(1.00) During-normal power operation the NNI System supplies a Unit Tave f signal'to the ICS, DESCRIBE what will cause the Auto / Manual Transfer 1 l switch to' automatically transfer to loop Tave. (1.0) ANSVER 6.15 (1.00) A RCS Flov error of'10% between loops vill cause the auto transfer of Tave'to the loop with_the highest flow. (1.0) TECO PVR-OLC-035 DAVIS BESSE TRAINING INFORMATION MANUAL VOL. 5 NNI 016000K101 ...(KA'S) RESPONSE 6.15 Recommend accepting for full credit loss of NNI power supply, which will cause a transfer. REFERENCE PROVIDED AB 1203.41 Loss of NNI Power Page 17 of 31 l l

l QUESTION 6.18 .(1.50) 1. What determines control bleed off flow under normal conditions? (0.5) 2. VHAT recent modification was made to increase the amount of fluid between'the RCP seals to reduce seal failure? (0.5) 3. What one specific failure mechanism did this modification address? i (0.5) ANSVER 6.18 (1.50) 1. The size of the staging coils, (capillary tubes) 2. Cut slots in the face of the Titanium carbide Seal face rotting element (this' increased the gap between the faces due.to the sheer -forces of the fluid ie. hydrodynamic forces, between the faces) 3. (To increase the cooling flow between faces and to increase the gap between the faces,) to reduce (the brake horse power) heat stress cracking of the faces. REFERENCE SP 1103.06 PVR-OLC-033 DAVIS BESSE TRAINING INFORMATION MANUAL VOL. 1 RCP PUMP & HOTOR 003000K602 ...(KA'S) RESPONSE 6.18 Parts 2 and 3 are double jeopardy. If candidate does not know Part 2, he can not get Part 3. Recommend Part 3 be deleted. REFERENCE PROVIDED NUREG 1021, ES 202 l Page 18 of 31 j j I L_i___- ___ -- )

j.. ' \\, QUESTION 6.21 (2.00) l 4 When the Component Cooling Surge Tank Level decreases to 35 inches, level switches will close or block the opening of four.(4) valves or group:of. valves in the Component' Cooling Vater System. LIST four (4) of'the six (6) major components that lose CCV flow-upon lov CCV Sur.ge Tank level. I ANSVER '6.21 (2.00) 1. Control Roj Drive Mech. 2. Reactor Coolant Pumps 3, Reactor Coolant Motors o i 4. Make Up Pumps 5. Emerg. Instr. Air Compressor 6. Letdown Coolers j REFERENCE l GOP-OLC-001 Attachment 31 I AB 1203 31 k/a R0 3.6 SR0 3.5 000026G010 008000K301 ...(KA'S) RESPONSE 6.21 Question asks for four (4) major components. Definition of " major" is difficult. Request that all Joads which are isolated be considered in addition to the six (6) listed in key. 35" vill isolate the Containment Header, Control Rod Drive Header, and the Make-up Pump Header. See reference for list of loads. REFERENCE PROVIDCD SP 1104.12 Page 19 of 31 1

7-. L.. i QUESTION 7.01-(2.00) i l EP 1202.01, LACK OF ADEQUATE SUBC00 LING MARGIN has you secure RCPs l vithin two (2) minutes of loss of subcooling margin. What are the BASES for that action include the bases for the two (2)' minutes?' ' ANSWER 7.01: (2.00) j RCPs must be tripped within two minutes after losing SCM to prevent the RCS'from reaching-a significant void fraction (0.5) such that the core vould be uncoveted if the RCPs were tripped at a later time. (1.0) The:two minutes is based upon the small break analysis, the earliest that a significant (70%) void fraction vould occur, (0.5) t REFERENCE I GOP-0LC-003 TECHNICAL BASES DOCUMENT pg IV.A-2 000074EK30 ...(KA'S) RESPONSE 7.01 Point breakdown in answer key appears inconsistent with question asked. Entire question is answered in first sentence of answer key. Recommend question be scored accordingly. l I i Page 20 of 31 a

0UESTION 7.03 (2.00) 1. STATE the limits given in AB 1203.40, Steam Generator Tube Leak for OTSG Tube to Shell delta T. for the following: 1. Normal tensile tube-to-shell delta T (tubes colder) (0.5) 2. Compressive tube-to-shell Delta T (shell colder) (0.5) 2. State how to determine (include instrumentation): 1 1. Tensile tube-to-shell Delta T (0.5) 2. Compressive tube-to-shell Delta T (0.5) ANSVER 7.03 (2.00) 1. 1. 100 deg. F 2. 50 deg.-F 2. 1. Average SG Shell Temperature - T cold (tensile) 2. T hot - Average SG Shell Temperature (compressive) REFERENCE AB 1203.40 step 3.21 TECIINICAL BASES DOCUMENT SEC. III.G-19 G0P-OLC-003 035010G005 ...(KA'S) RESPONSE 7.03 Part 2 recommend accepting the use of RCS average temp. and OTSG average shell temp. as an alternate method of determining tube to shell delta Ts for full credit, as contained in Plant Shutdown and Cooldown PP 1102.10. REFERENCE PROVIDED PP 1102.10 L i Page 21 of 31 __-__________a

j n f. l f 00ESTION 8.02 2.50) J ( i L List five (5) conditions which would result in declaring a movable l control assembly inoperative in accordance with Tech, Specs. Do not 1 include possible causes for the condition in your answer. f ANSVER 8.02 (2.50) (any 5 of the 6) 1. Cannot be moved 2. Ca.not be locked 3. Control rod is misaligned with its group average by more'than 9 l inches 4. Control rod does not meet the exercise requirements 5. Control rod does not meet the rod trip insertion tiens 6. Control rod does not meet the rod program verification (5 @ 0.5 ea.) REFERE14CE TECO T.S.'3.1.3 sec. k/a R0 3.4 SR0 3.9 001050G011 ...(KA'S) 4 RESPONSE 6.-02 Answer 1, the question states that the assembly is movable, and yet Answer 1 says it cannot be moved. Recommend accepting four (4) ansvers for full credit. Answer 2, rod that cannot be located would imply the rod position is inoperable. Recommend necepting this response for credit. Answer 3, request that alternate response rod misaligned by more than 6.5% vith its group average be considered for full credit. Request that other alternate responses be considered for credit, including: 1. Failure to perform a Surveillance Test within the specified time period. 2. A safety rod that cannot be fully withdrawn must be declared inoperable. REFERENCE PROVIDED Tech. Specs. 3.1.3.1, 3.1.3.3, 4.0.3, 3.1.3.5, 3.1.3.7, 3.1.3.4 Page 22 of 31 i 1

QUESTION 8.03 (1.00) Per AD 1850.04, Post Accident Radiological Sampling and Analysis, which of the following is the responsibility of the Emergency Plant Manager? (Choose.one) Responsibilities: a. Contacting the Emergency Control officer at NRC. b. Authorizing that a PASS sample be taken. c. Coordination of sample movement to and from the analysis facility. d. Determination of maximum allovable personnel dose to obtain the sample. ANSVER 8.03 (1.00)

1. b REFERENCE TECO AD 1850.04 "Pos t Accident radiological Sampling and Analysis" 000076EK30

.../KA'S) RESPONSE 8.03 Request question be deleted. This question tests candidates on the actions of the Emergency Plant Manager providing authorization to C&HP personnel. The SR0 candidates will fill neither of these positions. Page 23 of 31 j

QUESTION 8.05 (1.50) 1. .If a Safety Limit is viola'.ed, what 2 notifications are. required i as given in AD 1839.00, Sta:;on Operations, after completion of j T.S. required actions? 2. Who authorizes resumption of operation? ANSVER 8.05 (1.50) l 1. 1. Notify Assistant Plant Manager-0perations (0.5) 2. Notify NRC (within one hour) (0,5) 2. Ensure operation -is not resumed until approved by.the K<C (0.5) ~ REFERENCE ADH-SRO-001.01 I.1 AD 1839.00 k/a R0 2.5 SRO 3.4 194001A103 ...(KA'S) RESPONSE 8.05' Request you consider the following alternate response from Tech. Specs., Section 6.7: 1. Notify Senior Vice President, Nuclear 2. Notify Company Nuclear Review Board REFERENCE PROVIDED Tech. Specs., Section o.7 i i l 1 Page 24 of 31 L.________

i l QUESTION 8.06 (1.50) LIST the documents relating to Surveillance Testing the SS and Assistant SS review prior to turnover. ANSVER 8.06 (1.50) 1. ST Schedule (0.5) 2. ST Alert Report (0.5) 3. ~ Critical ST Report (0.5) REFERENCE ADM-SR0-004.01 B.1' AD 1839.05' - k/a R0 2.5 SRO 3.4 194001A103 ...(KA'S)~ RESPONSE 8.06 Recommend considering a word description of these three (3) schedules -l 1 1 1 1 l l [ I [. l 4 Page 25 of 31 j I J

r, QUESTION 8.07 (1.50)- I g l. According to AD 1803.00,- Safety *agging, restoration to service'shall not begin until what'three actions are performed? ANSVER 8.07 '(1. 50) 1. Work'has been completed. (0.5) 2. All personnel holding. clearance.have released clearance. . (0,5) j 3. The SS/ Assistant SS have determined that the equipment is ready i for service. (0.5) REFERENCE ADH-SRO-005.01 TECO AD 1803.00 i k/a R0 2.5 SRO 3.4 194001A102 ...(KA'S) RESPONSE 8.07 Request alternate responses'be considered for full credit: 1. Determin'ng required tests or inspections. 2. Safety tags have been removed. 3. Lineups required for return to service. 4. Electrical lineups. 5. System filled and vented. 6. Maintenance personnel present for post-maintenance' running of equipment. 7. Satisfactory disposition of PCAQ or NCRs. 8. MVO signed off. REFERENCE PROVIDED AD 1839.00 Station Operations AD 1844.02 Control of Vork l l i Page 26 of 31

w~ ';i, . l p. i*7, 3 ., y OUESTION 8.08" '(1.50)- l J.y. n / ccording.to.AD 1823.00;-Jumper and Lifted Vire CoM rol, b; 1. Installation of a jumper on Critical'Lystems/ equipment requaies l-- l . independent verification by whont? + . 2. Vho determines whethe'r a system is critical or not when a-jumper is to be installed? ANSWER 8.08 (1.50)- ~ 1. A qualified Journeyman'or above. -(0.75)' . 2. > The DBTS'(Davis-Desse Tagging Supervisor) (0.75) REFERENCE" ADM-SRO-006.-B.1 . TECO AD.1823.00.17 - k/a R0 2.5 : SRO 3.4' 194001A110 ...(KA'S)- RESPONSE 8.08' Request accepting Shift Supervisor as an' alternate response for full credit for Part 2. REFERENCE PROVIDED-AD 1823.00 Jumper and' Lifted Vire Control i I 11 'l l i i l Page 27 of 31 i i l _i i- -.- Li_____.___.._.____._ ___J_',

i ) QUESTION 8.12' (1.00) FILL IN Tile BLANK Tec. Spec. 4.0.2 States: f Each Surveillance Requirement shall be pr.rformed within the specified time interval with: a. A maximum allowable extension not to exceed of the. surveillance interval, and b. A total maximum combined interval time for any 3 consecutive . tests not to exceed times the specified surveillance ). interval. ANSVER 8.12 (1.00) a. 25% (0.5) b. 3.25 '(0.5) REFERENCE-TECO TECilNICAL SPECIFICATION 3/4.0.2 k/a R0 2.5 SR0 3.4 194001A103 ...(KA'S) l RESPONSE 8.12 i Request you also accept 1.25 of interval for Part a. as an alternate response for full credit. l i Pa,e 28 of 31 i

QUESTION 8.15 (1.50) Per D-B Emergency Plan Sec 6.5.1, Emergency Personnel Esposure: 1. What are the exposure dose guidelines for: 1. Lifesaving action 2. Corrective action l 2. The Shift Supervisor' vill initially have the authority to permit the Emergency Exposures. Once the Technical Support Center is activated, the responsibility is shifted to VHOM? ANSVER 8.15 (1.50) 1. 1. 75 rem (0.5) 2. 25 rem (0.5) 2. Emergency Plant Manager (0.5) REFERENCE D-B Emergency Pln Sec 6.5.1 k/a R0 2.5 SRO 3.4 194001A116 194001K103 ...(KA'S) ) RESPONSE 8.15 Recommend accepting Emergency Director as an alternate response for full credit to Part 2. The Emergency Director initially has the responsibility and can delegate it to the Emergency Plant Manager. REFERENCE PROVIDED Davis-Besse Emergency Plan i i Page 29 of 31 3 1

-l QUESTION'8.16' ( 1. 50'). Per D-B Emergency Plan LIST the three (3) designated areas where the emergency teams can-be assembled and equipped. ANSVER 8.16 (1.50) 1. Turbine Deck (0.5) 2. Health Physics Monitor Room (0.5) '3. Radiological Testing Laboratory (0,5) i REFERENCE D-B Emergency Plan'Sec 6.3.2 k/a R0'2.5 SRO 3.4 194001A116 ...(KA'S)- RESPONSE 8.16 Request Operation Support Center be considered as an additional response for: credit. REFERENCE PROVIDED Davis-Besse Emergency Plan I i l I 1 I Page 30 of 31 1 1 m._._.____

s. 5 jl.. r' ti " I

GENERAL CONCERNS ON ' SRO. EXAM '

a (Category'7Lof the'SR0 exam' contained.five (5) q'uestions (no. l', : 3 ' 4,. 5,farid 11) vorth 8.5 points, which were-derived from the Technical Bases Document. This represents approximately 35% of,the: total; points "for this' category. .This document was produced bylthe B&W Owners Group.and wasLdeveloped primar.11y to establish a. technical bases format.that/provides an 1 i efficient ~vehicleLfor document maintenance and: periodic updates to", l address new. issues!and operational. methods on a generic bases. c i .This: document isinot used as training material and was' not submitted with.the initial reference request. Ve recommend Lthat in fthe future, this document notL be.used 'for the preparation of operator exams. f REFERENCE ? Technical' Bases Document, Part'I, page'I.1-m l-CJ I-Page 31 of 31 Li.. - -- -- --- i

i U. S. NUCLEAR REGULATORY COMMISSION j- ' REACTOR OPERATOR LICENSE EXAMINATION FACILITY: DAVIS-BESSE 1 REACTOR TYPE: PWR-B&W177 DATE ADMINISTERED: 87 /00.Z17 EXAMINER: SPENCER. M. l l CANDIDATE l 1 INSIBUCTIONS TO CAHDIDAIE1. 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 l crade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts. i % OF CATEGORY % OF CANDIDATE'S CATEGORY t VALUE_ TOTAL SCORE VALUE CATEGORY j 24.50 24 b 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW l ~lG. bo 25.$ _LM 2. PLANT DESIGN INCLUDING SAFETY l AND EMERGENCY SYSTEMS so ,_g6.50

26. h1 3.

INSTRUMENTS AND CONTROLS 'k_ i $ 4. PROCEDURES - NORMAL, ABNORMAL, CGRGENCY AND RADIOLOGICAL CONTROL lO o. 00 t h e._. Totals Final Grade All work done on this examination is my own. Ihavene(tgey,.given nor received aid. Candidate's Signature

l I NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS l Duringthe. administration of this examination the following rules apply: ' 1. Cheating on the examination meano an automatic denial of your application and could result in more severe penalties. 2. Restroom trips are to be limited and only one candidate at a time may ' leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating. 3. Use black ink or. dark pencil only to facilitate legible reproductions. 4. Print your name in the blank provided on the cover sheet of the examination. 5. Fill in the date on the cover sheet of the examination (if necessary). i 6. Use only the paper provided for answers. J '7. Print your name in the upper right-hand corner of the first page of-each section of the answer sheet. 8. Consecutively number each answer sheet, write "End of Category __" as j appropriate, start each category on a new page, write only on one side j of the paper, and write "Last Page" on the last answer sheet. 9. Number each answer as to category and number, for example, 1.4, 6.3.

10. Skip at least three lines between each answer.

11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.

12. Use abbreviations only if they are commonly used in facility literature.

I 13, The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.

14. Show all calculations, methods, or assumptions used to obtain an answer t

to mathematical problems whether indicated in the question or not.

15. Partial credit may be given.

Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

16. If parts of the examination are not clear as to intent, ask questions of the examiner only.

17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in coinpleting the examination.

This must be done after the examination has been completed. l e o

.; :r e18;rWhen you complete 1your examination, you chall: j l L ' c '.' : Assemble your. examination'as follows: -(1);' Exam questions on top. .(2) Exam cids - figures, tables, etc. (3); Answer pages including figures which are part of the answer; t b. Turn.in your copy of the examination and all pages used to answer I the' examination questions. . Turn in.all scrap paper and the balance of the paper.that you did i c. not use for answering the questions. ~d. Leave theLexamination area, as defined.by the examiner. If after. leaving, you are found in this area while the examination is still in: progress, your license may be denied or revoked. t I j l 1

til PRINCIPLE 800F NUCLEAREPOWER PLANT OPERATION.-

Paga

'4' \\ IHEBHODYNAMICSr HEAT' TRANSFER AND FLUID FLOW. g *, ' ' l ': ' QUESTION fl. 0'1!

(1.00) l Define the. term Departure from Nucleate' Boiling Ratio.

QUESTION 1.02 ' ( 1. 00 ): How'can a decreasing pressurizer level affect natural circulation flow? l NO RCS leakage is present. QUESTION-1.03 (2.00) List the'FOUR heat transfer regions in an OSTG and indicate how each region changes (INCREASE, DECREASE OR REMAINS THE SAME) as power l I is increased from 20 to 100%. QUESTION 1.04 (1.00) What'are FOUR engineering practices used to minimize waterhammer by.the operator? j QUESTION 1.05 (1.50) Explain.HOW and TWO reasons WHY natural circulation flow would be affected by the operator manually raising the OTSG level above the low level limits. ASSUME natural circulation flow was stable before changing the level. 1 1 l l \\ (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) l J

3, 8 - i w Y-15 I 4 Paco _5: PRINCIFLES OF NUCLEAR-POWER PLANT OPERATION. N E AT TRANSFER-AND FLUID FLOW ] d 5

- $ l

) J QUESTION- ; 1'. 0 6 .(2.00)

Indic' ate HOW each'of the following items are affected during a-RC3 heatup.

. Answer with' INCREASE, DECREASE 1or NO CHANGE. ') A. .RCP head. { l l B. RCP power. C.. RCS~ volumetric flowrate, q JD. 'RCS mass flowrate. ) QUESTION 1.07 (1.00)- Which one of1the following statements.about pump Net Posi+' m "oction Head ~(NPSH)Lis CORRECT? A. NPSH its the amount by which the saturation pressure is sz. ster than the auction pressure for.the water being pumped. i .B. When a pump-is started, the NPSH will decrease by the amount of the pressure drop in the suction piping. C. NPSH is essential for operation of centrifugal pumps but not for' positive displacement pumps. D. NPSH can be calculated by subtracting the suction pressure from the discharge pressure. QUESTION-1.08 (1.00) Why should the APSR position remain unchanged, following a reactor trip? QUESTION 1.09 (2.50) List.FIVE of SIX factors which are considered in shutdown margin when the reactor is in'a shutdown mode. i l l l 1 (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

Y -1' PRINCIPLES OF NUCLEAR' POWER PLANT OPERATION.- Paga 6

IHERM0 DYNAMICS,-HEAT _IRANSFER AND FLUID FLCW-

+ t QUESTION 1.10 (1.00) Indicate HOW (LESS NEGATIVE, MORE NEGATIVE cnr NO CHANGE) differential ] v boron worth changes with.an increase in'each of the following: -A. . Boron concentration i B. Moderator temperature 1 i QUESTION. 1.11 (1.00) Which one of the following is CORRECT concerning Moderator Temperature-J Coefficient (MC)? l A. MTC is not-permitted by' Technical Specification to be h . posit.tve for any plant operating mode. B. MTC becomes less negative as fuel depletion occurs. C. MTC causes axial flux distribution to be tilted toward the top of the core. D. MTC becomes more negative as +.emperature increases due to non-linear expansion of the n derator. QUESTION 1.12 (1.00) A. What is the approximate 50% BOL equilibrium value for Xenon? 1 B. What is.the approximate peak BOL Xenon value for trip from 100%? s 4 r l l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) l a

1.

PRINCIPLES'OF. NUCLEAR POWER PLANT OPERATION, Pago 7

THERMODYJiAtilGS, HEAT TRAN3EER AND FLUID FLOW j

I a sJ 1 l l QUESTION 1.13 (1.00) y During a:startup it was determined that Keff '.es equal to 0.9 when.the i Source Range (SR) instru:acnt tras reading 50 cps. What would the source { range instrument be reading if rods were withdrawn to bring Keff equal. .l to 0.957 Assume BOL conditions. A. 75 cps B. 100 cps J d

C.

125'eps D. 150 cps ? ll QUESTION- ~ 1.14. (1.00) Explain HOW and WHY the Source Range indications would behave during d core voiding from-a LOCA condition. I LQUESTION 1.15-(1.50) ') Briefly explain HOW the power coefficient changos over core life, I (MORE NEG., LESS NEG., or NO CHANGE) and WHY. l l QUESTION 1.16 (2.00) 1 HOW would the actual critical rod position vary (HIGHER, LOWER, or THE i SAME from the estimated critical rod position (ECP) for EACH of the j following situations. Include a BRIEF explanation WHY. J A. After a trip from 100% power, an ECP is calculated using zero k xenon reactivity for a startup 8 hours after the trip.. 1 B. The actual boron concentration is 100 ppm lower than the value used for the ECP. ] 'C. -l20 EFPD is used in the ECP instead of the actual 200 EFPD. D. The source count rate has decreased from 20 cps to 10 cps ~ between the time the ECP was calculated and the beginning of the startup. l l-(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) j l i L-

w '

eK;; 3 M ili. PRINCIPLES OF'~ NUCLEAR-POWER PLANT OPEEN[1Qth.

Pass!'8 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW

-{ 4 r <., I., ' k 1. ',' f [ t 3 [ QUESTION:

1.17, (1.00);

Why are SelffPowered= Neutron Detectors (BPNDs) NOT used.for reactor q ~ control purposes?' 1

}

i.g., ~ \\ 5,.' QUESTION : 13.18 (1.00)' i .a L.

The ratio of both'Pu-239 and Pu-240 atoms to U-235. atoms changes over-3 J

the core life.' Which one of the pairs of-parameters below are most f - affected by.this change? j q ~ A. ' Moderator-temperature coefficient and doppler coefficient B. Doppler'. coefficient and beta ' C-- Beta'and thermal neutron' diffusion length. ~ Di Thermal neutron diffusion length and moderator temperature h . coefficient.' q r .i 1 i 1 1 l i l I (***** CATEGORY 1 CONTINUED'ON NEXT PAGE *****) 1 k m--a.rnu-w---_- -, - - _ - - - - - -,- a n

i l ,gl. PRIHCIPLES OF NUCLEAR POWER PLANT OPERATIONc Poga 9 { TEEBMODYNAMICS. UEAT TRANSFER AED FLUID FLOW i QUESTION 1.19 (1,00) Which THREE of the following will have an effect on the shape of a 1/M plot during fuel loading? 1. The location of the neutron sources in the core. ) 2. The strength of the neutron sources in the core. 1 3. The. location of the neutron detectors around the core. 4. The order of placement of fuel assemblies provided the proper enrichments are placed in their proper location. 5. The flow rate o the Decay Heat Removal system supplied to the reactor vessel. The correct conditions are: I l A, 1, 3, 5 B. 1, 3, 4 C. 2, 3, 5 i .D. 1, 2, 4 l I i MASTER C0)Y 1 (***** END OF CATEGORY 1 *****) -~

2; PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Paga 10-SYSTEMS' k. QUESTION-2.01. (0.75) 'The reactor coolant pumps employ a three seal package. List the typical pressures expected to be seen in the high pressure cavity side of each. seal:during normal power operation.- [ seal #1; seal #2; seal #3) j i QUESTION 2.02 (2.25) Each reactor coolant pump motor is equiped with an Anti-Reversing I Device. List THREE reasons reverse rotation is not desired and briefly EXPLAIN each reason. QUEGTION 2.03 (1.00) l -l Which of the'below best describes the location of the connection from the service water system to the aux, feed system? A. upstream of aux, feed pump suction valvo FW 786 ( FW 790 ) l and downstream of Fire Protection system inlet valve ( FP 28 ). 1 B. upstream of aux, feed pump suction valve FW 786 ( FW 790 ) and downstream of supply from the condensate storage tanks ( valve CD 170 ) C. upstream of aux, feed pump and downstream of valve FW 786 ( FW 790 ). D. upstream of the Fire Protection system inlet ( valve FP 28 ) and downstream of the supply from the condensate storage tanks. ( valve CD 790 ) i (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) l l 1 a

- f.- -2. PLANT-DESIGN INCLUDING SAFETY AND EMERGENCY Pega'11 SYSTEMS I i i QUESTION 2.04. (1.00) .The below question relates to the-Electrohydraulic Control System. Select from the descriptions in Column.B the item er device listed in Column A.- COLUMN.A-COLUMN B 1. Accumulators a. maintains fluid supply during testing of.the 2. Mechanical. Trip Valve turbine overspeed device 3. Haster Trip Solenoid b. provide reserve fluid during transient flow 4. Lockout Valve conditions c. directs fluid for operation of the main stop and intermediate stop valve control Pacs d. three way valve mechanically connected to manual trip lever. l QUESTION 2.05 (1.50) List the THREE electrical power supplies to the Electrohydraulic ] Control System. i l QUESTION 2.06 (2.00) The Station Air Compressors have several distinct differences. List FOUR of the instrument enc'. control differences. 1 i 1 (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) i c l

-l . 1 _iPLANT DESIGN' INCLUDING SAFETY AND EMERGENCY' Paco 12 SYETEMS- ~ ~ TQUESTION

2. 0 7 '

(2.00)' . Match the' location ~of the system interconnection with the Reactor Coolant System'in Column.B with the interconnecting system _ listed in -Column A. COLUMN A COLUMN B 1. Make-up. System a. taps off of 2-1 hot leg 2. Decay Heat System drop line b. discharge of each RCP-3. Pressurizer Spray line c.- discharge of 2-1 RCP 4. 'HPI tap d. discharge of 2-2 RCP QUESTION 2.08 .(2,00) i List the FOUR major differences in the design / construction of the control rod' assemblies and the axial power shaping rod assemblies. l QUESTION 2.09 (1.00) List FOUR sources available to the Borated Water Storage Tank. QUESTION 2.10 (1.00) Why is flow through orifices R06A and R06B of the High Pressure j Injection System not required during the " piggyback" mode" l ( Two reasons required for full credit ) f i 1 r (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) q i l 1 1 1

U x J. ~ PLANT DESIGN' INCLUDING SAFETY AND EMERGENCY Pega 13, i

SYSTEMS, l

8 ) I I UESTION 2.11 '(1.00) j Q Select the bestidescription of the location of.the Makeup pump recirculation line.to makeup system. A. between the seal injection filters and the manual-isolation valve. (19) B. l returns directly to.the makeup tank O. between the letdown flow indicator FE-7 and the demineralizers i D. between the seal' return header and the seal return-coolers j l QUESTION' 2'12 (1.00) l l List'TWO reasons for the spray valve bypass (RC-49) being ] throttled at 1.25 gpm. i 1 QUESTION 2.13 (1.00) l i List FOUR of.the five loads' supplied by ESSN LINE 1. of l the' Component Cooling Water System during normal conditions.

QUESTION 2.14 (2.00)'

A.. List the TWO purposes of the Emergency Ventilation System. (Setpoints or values not required; areas or rooms ARE required) J B. List TWO systems from which the Emergency Ventilation System has the capability of taking a suction. j -l l. 1 I l l l (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) l ) l ___-_-_____-__--____---___-__Q

. 2. PLANT DESIGN INCLUDlHQ_g/JETY AND EMERGENCY Pc30 14 SYSTEMS 4 QUESTION-2.15 .(1.00) List two reasons for the " piggyback" mode of operation of the HPI system. So QUESTION ~ 2.16 ( 1.W List the normal power supply by " bus" identification for each of the below: 1. .RCP 1-1 2. RCP 2-2 3. Station air compressor 1-1 4. Emergency. air compressor 5. .HPI pump 1-2 D::;7 heat N...; r MP% udM n '9 S 7. CCW pump 1-1 QUESTION 2.17 (2.00) A. Manual Deluge System is one of the six different types of fire suppression systems used at Davis-Besse. List the remaining five fire suppression systems. (1.25) B. List the three sources of water for the fire protection systems. (0.75) (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)

L', 2.lIf' ANT DESIGN INCLUDING SAFETY AND EMERGENCY Paga 15 k p,

QUESTION 2.18 (1.00)

The Core Flood Tanks can be filled using three different methods. T List TWO of these methods. MAHR C0P ' (***** END OF CATEGORY 2 *****) )

3; INSTRUMENTSLAND CONTROLS Paso 16 v ,o QUESTION. 3.01 (2.00) The Integrated Control System shifts to track mode when any of its stations / control panels are placed in manual. List FOUR other conditions or events causing the ICS to shift to track. l QUESTION 3.02 (2.00) State FIVE of the seven RCP starting interlocks. Setpoints are NOT required. QUESTION 3.03 (2.00) State the FOUR ICS runbacks, their rates and their limits. L QUESTION 3.04 .(1.50) What THREE conditions will prevent the feedwater section of the ICS from taking T-ave control when reactor demand is taken to hand? QUESTION 3.05 (2.00) What FOUR trips are active on a Loss of Voltage or SA start condition on the Emergency Diesel Generator? (Setpoints not required) QUESTION 3.06 (1.50) What FIVE conditions will cause the Steam and Feed Rupture Control System (SFRCS) to automatically initiate Auxiliary Feedwater? (Setpoints are not required). QUESTION' 3.07 (1.00) What must be done on the control board to open either Auxiliary, Feed Pump Turbine steam supply valve, if a low AFP suction pressure { condition exist? (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

Sl ' INSTRUMENTS'AND CONTROLS Paco 17 QUESTION 3.08 (1.00) What are the automatic actions that occur in the event the setpoint is' reached on RE-1822 A or B (Waste gas)? QUESTION 3.00-(1.00) Which one of the following statements concerning the Control Rod Drive Position Indication System is CORRECT? A. The zero-(0) percent switch is located 1.5 inches above the in-limit' switch. B. .The 100 percent switch is located 1.5 inches above the out-limit' switch, .C. The first rod in any group to reach the 100 percent switch will stop further travel of all rods in that group. D. When. actuated, the out-limit switch will generate an out-inhibit condition on the Diamond Panel. QUESTION 3.10 (2.00) l l NAME the TWO cross limits associated with the ICS and BRIEFLY describe each cross limit. Include in your discussion, the conditions under which the limit will be in effect and the demand signal (s) which will be modified by the limit (including direction of change). l l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) 9

3. INSTRUMENTS AND CONTROLE 'Paga 18 V l d QUESTION 3.11 (1.00) Which'ONE of the statements below is CORRECT concerning the Makeup and Purification System: 1 A. The block orifice has two bypasses, both of which are remotely operated from the control room. -B. The letdown line connections to the Decay Heat Removal System are prior to the letdown filters and after the block orifice. C. A temperature element on the letdown line alarms at 130 degrees F and closes the letdown cooler outlet valves (MU2A and MUS) at 135 degrees F to protect the letdown coolers. D. The makeup demineralizers may be only operated in parallel. QUESTION 3.12 (1.00) Which one of the following RPS trips is NOT bypassed when the RPS is in " Shutdown Bypass"? A. Low Pressure B. High Flux 0. Flux / Delta Flux / Flow D. Variable Pressure Temperature (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

3. INSTRUMENTS AND CONTROLS Pega 19 I t QUESTION 3.13 (2.00) The turbine bypass valve may be biased by a O psig, a 50 psig, or a 145 psig signal. Indicate which one of these biases apply to each of the four situations below. A. The reactor and turbine are not tripped. the turbine bypass valves are closed, and header pressure deviation is less than 10 psig. B. The reactor is tripped as indicated by a TRIP CONF light on the Diamond Panel. C. The reactor and turbine are not tripped, all turbine bypass valves are NOT closed and ULD is greater than >(7 ona percent. 1 D. The reactor is not tripped, and the main turbine is tripped. QUESTION 3.14 (1.50) How would the plant respond if the Thot input to the ICS failed low while the plant was at 100% power with the unit load in full automatic with no operator action? Be brief in your explanation. QUESTION 3.15 (1.50) The Main Feedpump's speed is controlled by the ICS to maintain a constant differential pressure across the Feedwater control valves. How does the ICS anticipate a drop in the differential pressure when there is an increase in the demand for flow? QUESTION 3.16 (1.00) What are the TWO types of Rod Position Indication and briefly state how is each sensed? l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

3. INEIRUKENTS AND CONTROLS Pogo 2:0 I . QUESTION 3.17 (2.50) I A.. What FOUR actions will take place if a " Bus Lockout" occurs on the C1 essential bus? (2.0) l B. What action (s) occur as a result of depressing the " Lockout Reset" button? (0.5) l l i l l NASIR C?Y (***** END OF CATEGORY 3 *****) w_-__.

l 4. ' PROCEDURES -- NQBtML. ABNORMAL EMERGENCY Paga 21 i AND RADIOLQQICAL CONTROL l ' QUESTION. 4.01 (1.00) List FOUR of the six symptoms or entry conditions for Emergency Procedure EP 1202.01, 1 QUESTION 4.02 (1.00) According to Safety Tagging AD 1803.00, restoration to' service i shall not begin until what THREE items are completed? QUESTION 4.03 (1.00) List'FOUR'of the seven Reactor Operator responsibilities as identified 'in AD 1839.00,_ Station Operations. QUESTION 4.04 (1.50) List SIX different types or SIX different examples of information I i required to be entered into the Reactor Operator's Log per AD 1839.00. l I J l I \\ p l (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) I L_ _ _ _ -_ -

4 PROCEDURES - NORMAL. ABNQBti&L. EMERGENCY Pcgs 22 AND RADIOLOGICAL CONTROL J i QUESTION. 4 '. 0 5 - (2.00) j 1. Complete the blanks in the below peregraphs. l Any~ individual should not be permitted to work more than a-hours straight (excluding shift turnover time). An~ individual should not be permitted to work more than b hours in any c period, nor more than __d in any 48 hour I period, nor more than .e hours in any seven day period (all excluding shift turnover time). A break of at least __f hours should be allowed between l work periods (including shift turnover time). (NOTE each blank is worth 0.25 pt.) I 2. Any deviation past these restrictions will be approved by whom? (name or title of minimum level management) l QUESTION 4.06 (1.50) HP 1601.01.10, Guide and Limits for Exposure to Radiation places three limits or guides on quarterly and annual exposures. 1. What is the working quarterly guide limit? 2. What is the working quarterly maximum allowable limit? 3. What is the working annual guide limit? 1 QUESTION 4.07 (2.00) j SP 1103.04.8, Boron Concentration Control, list several requirements to stop boration or deboration. List FOUR of these requirements. (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

4. PROCEDURES - NORMAL ~. ABNORMAL. EMEBGEHfd Pago 23 -AND RADIOLOGICAL CONTROL h QUESTION 4.08 (1.50) EP 1202.01 RPS, SFAS, SFRCS, TRIP or SG TUBE RUPTURE, Section 6, Lack of Heat Transfer, requires availability of feedwater to be determined. List THREE criteria to determine feedwater availability. i QUESTION 4.09 .(2.00) List the FOUR actions required of the reactor operator per AB 1203.12, Control Room Evacuation, before leaving the control room. i QUESTION 4.10 .(1.00) EP 1202.01, Specific Rule 2, States the High Pressue Injection may be throttled and normal makeup flow established when two conditions are met. List these TWO conditions. QUESTION 4.11 (1.00) When tagging a valve closed, per AD1803.00, it is mandatory the valve be positively verified closed prior to hanging the tag. List TWO methods required to verify a manual valve is closed prior to hanging a tag. i (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pags 24 AE RADIOLOGICAL CONTROL L j-I QUESTION 4.12 (2.00) i A. Adequate subcooling margin exists when the T-sat meter indicates greater than or equal to which of the below? A. 10 degrees F B. 15 degrees F C. 20 degress F D. 35 degress F (0.5) B. According to the Emergency Procedure, EP 1202.01, Specific Rule 1, what THREE actions are required to be performed on the makeup system in the event of a loss of adequate subcooling margin? (1.5) QUESTION 4.13 (1.00) List the FOUR duties of the Primary and Secondary Reactor Operators. (EP 1202.01) QUESTION 4.14 (1.00) EP 1202.01.00, Section 4.1 Supplementary Actions for electrical power contains the below CAUTION statement: If a diesel generator fails to auto start, do not re-energize a 4160 volt bus with a makeup pump breaker closed and MU-19, Seal Injection, in AUTO as damage to the Reactor Coolant Pump seals could occur. Briefly explain this statement. QUESTION 4.15 (2.00) List the THREE conditions required to enable the ICS feed and bleed permissive. l l (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) t l I I I

%.<pa-' c h'ti.f 4; -PRdcRpl!BES'- NORMAL. ABNORMAL. EMERGENCY, Pega 25 - N MD" RADIOLOGICAL CONTROL s QUESTION 4.16 (1.50) 1. A reactor coolant pump must be tripped within what maximum i time interval if both seal injection and CCW are lost? 2. A reactor coolant pump must be tripped within what maximum time interval if ONLY CCW is lost and NOT restored? QUESTION 4.17 (1.00) 1. According to HP 1601.03.10, Health Physics Procedure, a pocket dosimeter should read less than what value before entering a i RACA7 2. According th HP 1601.03.10, Health Physics Procedure, a person shall exit a RACA when his pocket dosimeter reading exceeds what value? QUESTION 4.18 (1.00) 1. What approximate power level on the intermediate range instrument corresponds to the point of adding heat? (0.5) 2. What value should startup rate (SUR) be below prior to l the point of adding heat? (0.5) 1 NAS"R C?Y ~ (***** END OF CATEGORY 4 *****) (********** END OF EXAMINATION **********)

I' ,, E

• EQUATION SHEET s

\\ !f = ma 'v = s/t et rk (out)- ~ w = og s=vt+- hat ' Cycle efficiency = o Energy.(in) 2

E = aC a = (vg - v )/t g

~A C - A = AN KE'= lsev vf = v, + a t A = A,e PE = msh w = e/c A = in 2/tg = 0.693/tq- = v4P'

• h("II) * (g,,)(g )

-AE = 931am I4*'*b) Q = p,C AT g, 7,-Ix p ,, Q = UAAT g, 7,-px ~Pwr = W' m" -x/TVI, g I=I 10 l 5UR(t), g, y,37 P=P 10 e /T t HVL = 0.693/u P=P 'SUR = 26.06/T T = 1.44 DT SCR = S/(1 - K,gg) /A{f) CR, = S/(1 - K,gg,) p e SUR = 26 g 3( - K,gg)g = CR ( eff 2 7 '(1*/p ) + [(6 'o)/A,gg ] ~ 2 o 7,= t*/ (, _ p M = 1/(1 - K,gg) = Ca /CR t 0 T = (I - p)/ A,gg p g " (1 ~ Esff)OIII ~ Eeff)1 8 * (Esif"I)I eff " #eff aff /K SDM = (1 - K,gg)/K,gg [L*/TK,gg.] + [H/(1 + A,gg )] T 1* = 1 x 10 seconds ~ p= P = I4V/(3 x 1010) ~ A,fg = 0.1 seconds A E = No Idyy=1d22 WATER PARAMETERS Id =Id g 2 2 1 gal. = 8.345 lbm R/hr = (0.5 CE)/d g,,C,,,) 1 gal. = 3.78 liters R/hr = 6 CE/d (feet) l 1 ft = 7.48 gal. MISCELI.ANEOUS CONVERSIONS 3 10 Density = 62.4 lbm/ft 1 Curie = 3.7 x 10 dps Density = 1 gm/cm 1 kg = 2.21 lba 3 Heat of va;orization = 970 reu/lbm 1 hp = 2.54 x 10 BTU /hr 0 Heat of fusica = 144 Btu /lbm 1 Hw = 3.41 x 10 Btu /hr 1 Atm = 14.7 psi = 29.9 in. Ig. 1 Btu = 778 f t-lbf 2 1 ft. H O = 0.4333 lbf/in 1' inch = 2.54 cm 2 F = 9/5 C + 32 'c = 5/9 ( F - 32)

1. . PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Paga 26 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW i NEER C:)Y i ' ANSWER 1.01 (1.00) DNBR = heat flux required to reach DNB Critical heat flux or Actual heat flux Actual heat flux REFERENCE DAVIS-BESSE, HTR-OLC-005-00, Objective 3. 193004K110 ..(KA's) ANSWER 1.02 (1.00) The outeurge of the hot water can flash to steam in the hot leg and cause a loss of NC. (Void in candy cane) l REFERENCE DAVIS-BESSE Exam Question Bank #01-35 l DAVIS-BESSE, HTR-OLC-011-00, Objective 2 & 3 t (KA's) 193008K122 193008K121 e ANSWER 1.03 (2.00) Feedwater hea.ti.m Agreng/ Preheating.c INCREASE 1. cec.mv hw A 2. Nucleate boiling INCREASE 3. Film boiling REMAINS THE SAME or sh htl seWu 3 y 4. Superheat DECREASE REFERENCE, DAVIS-BESSE Exam Question Bank #01-29 l DAVIS-BESSE Training Information Manual, Vol 1, OSTG, pages 18-21 l l l N AS"R C:)Y I (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) l

ll ' PRINCIPLES OF NUCLEAR POWEB_P1hNT OPERATION.. Pese 27 ] IHEBtiQDlHAMICS. HEAT TRANSFER AND FLUID FLOW 4 1 i ANSWER 1.04 .(1.00) 1. Slow' opening of valves between voided system and full system. f 2. Proper. venting of components prior to starting. 3. Ensuring adequate levels in tanks, where these tanks provide a supply and/or a surge function. '\\ 4. Proper use of steam traps and vents.

I l

5. Following operating procedures. 4,. w.,,4,y we 54 tik 3 bd.n un'intJ *a (Any 4 @ 0.25 ea = 1.0) y, s4 M*..,g,,,,,. wi4L J os ekw 44 fot - ned REFERENCE DAVIS-BESSE Exam Question Bank #01-55 DAVIS-BESSE, HTR-OLC-022-00, Objective 4 061000K505 056000K503 039000K501 193006K104 ..(KA's) l ANSWER 1.05 (1.50) Raising the level would tend to increase natural circulation [0.5] by raising the effective OTSG thermal center, FEG8% ami thus increasing M.at source and heat sink'[0.5]and the h.eight differeOce between th h ue ..d heJ w to. b.w.en gg..,4 i nece., w d us.% A. e ere<. - e 9 REFER $$CE D DAVIS-BESSE Exam Question Bank #01-34 DAVIS-BESSE, HTR-OLC-011-00, Objectives 1 & 3 193008K123 ..(KA's) I ANSWER 1.06 (2.00) ) f A. NO C"?.MCE bu l B. DECREASE C. IMOREAE4 no % l D. DECREASE a (4 @ 0.5 ea,= 1 0) { 1 (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) l

L. Il ~PRINCIP M R'OF NUC M AR POWER PLANT OPERATION. . Pass 28 THERMODYNAMICS.' HEAT TRANSFER AND FLUID FLOW-r.. [- REFERENCE l-Generic: Centrifugal Pumps and System Hydraulics, Igor Karrassik. DAVIS-BESSE, HTR-OLC-019-00, pages 1-15 191004K105 191004K107 193006K101 ..(KA's) l ANSWER' 1.07 (1.00) B REFERENCE DAVIS-BESSE, HTT-SRO-013-00, pages 5 191004K101 191004K106 ..(KA's) 1. ANSWER-1.08 (1.00) They may insert positive reactivity depending on their position. REFERENCE -DAVIS-BESSE Exam Question Bank #01-106 192005K106 ..(KA's) ANSWER 1.09 (2.50) i 1. RCS temperature 2. Boron concentration 3. Xenon concentration l 4.

Samarium concentration 5.

Control rod position er b bi d werd 6. Fuel burnup based on gross thermal energy generation (3 @ 0.5 ea = 1 5) 5 (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) A

o1 PRINCIPLES OF NUCLEAR POWER PLANT' OPERATION. Paga 29 THERMODYNAMICS.-HEAT TRANSFER AND FLUID FLOW N. b JREFERENCE. DAVIS-BESSE Exam Question Bank #01-116a j ' DAVIS-BESSE, RTR-OLC-023-00 DAVIS-BESSE, Technical Specifications 4.1.1.11.e, page 3/4 1-2 192002K114 ..(KA's) s l ANSWER' 1.10 (1.00) A.. less negative- ~ (2 @ 0.5 ea = 1.0) t REFERENCE DAVIS-BESSE, RTR-OLC-017-00, page 5, Objective 2 192004K110 192004K109 ..(KA's) 'l i ANSWER 1.11 (1.00) D (1.0) i REFERENCE l l DAVIS-BESSE, RTR-OLC-011-00, pages 3-6, Objective 2 i 192004K107 ..(KA's) l ANSWER 1.12 (1.00) I. d 0.I A. 1644,% deltak/k + /- 4-4 % (0.5) 4.L O.E i B. ~bses deltak/k +/- 0,1% (0,5) -REFERENCE l DAVIS-BESSE, RTR-OLC-021-00, pages 5-6 l Objective RTR-OLC-021-00, #02 192006K102 ..(KA's) j j l i l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

i i E 1'. PRINCIPLES OF NUCLEAR _. POWER PLANT OPERATION. Pcga 30 THERMODYNAMICS. HEAT TRANSFER _AND FLUID FLOW 1 ANSWER l'.13 -(1.00) q B (1.0) REFERENCE ~ DAVIS-BESSE, RTR-OLC-024-00, Objective 1 192003K102 ..(KA's) ANSWER' 1.14 (1.00) Count rate would.significantly increase [0.5], With (erratic) oscillations due to the loss of neutron moderation andx_ i ve+4ure,before reaching the detectors. (1.0) j l REFEEENCE 1 DAVIS-BESSE Exam Question Bank #01-72 191002K117 ..(KA's) 1 l . ANSWER 1.15 (1.50) MORE NEGATIVE [0.5] -- due to the more negative [0.5]. component of moderator temperature coefficient [0.5]. (1.5) ] i REFERENCE DAVIS-BESSE Exam Question Bank #01-09 DAVIS-BESSE, RTR-OLC-015-00, Objective 2 192004K108 192004K113 ..(KA's) l l l I (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

~ 1' PRINCIELES OF NUCLEAR POWER PLANT OEERATION. Pag 3=31 ) THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW 1 ANSWER 1.16 (2.00) A. HIGHER [0.25].due to Xenon peak after the trip. [0.25] l B. LOWER [0.25] with less boron in the RCS the Control Rods 1will need ) to be inserted to compensate for the lower reactivityLof the boron.[0.25] C.. HIGHER'[0.25] due to'less excess reactivity remaining in the' core at 200 EFPD.[0.25] D. THE SAME.[0.25] since initial count rate.has'no effect on ECP (only power at which ECP occurs) [0.25] ] (Accept HIGHER if examinee states an assumption that negative reactivity has been added from some source to cause count rate to decrease.) (2.0) REFERENCE ' DAVIS-BESSE, RTR-OLC-015-00, pages 9-18 4 192008K104 ..(KA's) ANSWER 1.17 (1.00) The delay time of the Rhodium active material is too long for use. REFERENCE DAVIS-BESSE, Examination Question Bank, #1-10 015000K404 ..(KA's) ANSWER 1.18 (1.00) B (1.0) REFERENCE DAVIS-BESSE, RTR-OLC-013-00, pages 3-8 DAVIS-BESSE, RTR-OLC-010-00, pages 3-8 192003K107 192004K107 ..(KA's) { (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) I

l J. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Paga 32 i THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWER 1.19 (1.00) B. .(1.0) REFERENCE. General Physics' Academic Program for Nuclear Plant Personal; Vol II, ' Chapter 5 section A, pages 5-8 to 5-11 001000K518 ..(KA's) I ) i 1 i i l l l l kASTER CPY (***** END OF CATEGORY 1 *****) l __ _____ a

}., 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Paga 33 SYSTEMS l l I ANSWER 2.01 (0.75) l Seal #1 2155 psig Seal #2 1400 psig Seal #3 700 psig All pressures +/- 50 psig (0.75) l REFERENCE DAVIS-BESSE, TIM Vol 1, Reactor Coolant Pump and Motor, page 11. j l 003000A303 .. -( KA ' s ) j f ANSWER 2.02 (2.25) 1. CORE BYPASS FLOW (0.25) This would be due to discharge water, form the operating pump in that loop, flowing back through the non-operating pump discharge i piping and out the suction. A completely stopped RCP offers more resistance to flow than a "windmilling" RCP; therefore less core bypass flow is present if the RCP is completely j stopped. (0.50) { l 2. UPPER RADIAL BEARING DAMAGE (0.25) l This bearing is lubricated by an attached centrifugal pump. j If the attached lube oil pump is rotating backwards, it is pumping very little or no oil to this bearing. (0.50) 3. HIGH STARTING CURRENTS (0.25) j t If the RCP were allowed to rotate backwards and operator tried starting it, the pump motor would draw excessively high starting currents. (0.5) 1 (2.25) (SIMILAR WORDING ACCEPTABLE FOR CREDIT) REFERENCE DAVIS-BESSE, TIM Vol 1, Reactor Coolant Pump and Motor, page 13. 003000K608 ..(KA's) (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)

[j I 1 L2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pese 34 EXETEMS '^ ANSWER 2.03 ' ( 1. 00 )' i C-REFERENCE i DAVIS-BESSE, P & ID M-006D, Auxiliary Feedwater System j -061000K107 ..(KA's) I l -ANSWER 2.04 (1.00). . 1. = b .)

2. =

d-

3. =

c.

4. =

a .i i REFERENCE i DAVIS-BESSE,' TIM Vol 3, EHC, page 5. i 1 045050K603 ..(KA's) l ANSWER 2.05 (1.50) , j,c a g3 y o u. 1. Station Housepower,[ will accept 115 VAC 60 HZ ] 2. Permanent Magnetic Generator (PMG) 3. 125 VDC electrical distribution systemor 12Svoc pod D69 [ will accept station battery ] (3 @ 0.5 ea = 1.5) l l I (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)

f2. PLANT' DESIGN INCLUDING SAFETY AND EMERGENCY Pega 35 l P* SYSTEMS l l J c l~. l REFERENCE DAVIS-BESSE, TIM Vol 3, EHC, page 19. 045000K604 ..(KA's) l iANSWER. 2.06 (2.00) ) l 1. .On the-Auto Sentry Panel, the.first five: local alarms are the same. Sac 1-2 has alarm #6 for excessive vibration SAC.1-1 does not. i 2. The Auto Sentry Panel alarm #7 for low cooling water flow is for SAC 1-1 only. i 3. A start signal on SAC 1-1 closes its 480 VAC breaker, SAC 1-2 .I 480 VAC breaker must be closed for operation. A start signal will not close the breaker. 4. Pressing the "STOP" button in the control room will only stop SAC.1-1 as long as the button is held. To keep it stopped, the compressor switch.must be.placed in " LOCKOUT". SAC 1-2 can be stopped by pla'cing the control room switch in either "STOP" or " LOCKOUT". Low Jil pmsvec. 4r.'e sd?'N5 d8 f, sM a(c compe to.d.ad unload sdro,1b (4 @ 0.5 ea = 2.0) g, 5J a 3 R.Civut -REFERENCE DAVIS-BESSE, TIM Vol 8, Plant Air System, page 3. 079000K201 ..(KA's) ANSWER 2,07 (2.00)

1. =

c

2. =

a 3' = d 4. = b ( 4 @ 0.5 ea =,2.0 ) (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)

2. PLANT DESIGN INCLUDING SAFETY AND' ESRGENCY Pego.36; SYSTEMS 0, ^' REFERENCE-DAVIS-BESSE, PWR-OLC-034'.01, Reactor Coolant System, page 13 002000K109-002000K104' ~ ..(KA's) ANSWER 2.08 (2.00) '1. active length of-the silver-indium-cadmium ne on abso material is only the bottom 36 inches. 2. The female co Qngs of the CRA.and the RA have a slight dimensional difference. (This insures ea'cb.Qpe of can only'be coupled to the correct type of. h mechanism.) .3. The section of: tubing a e the ne on' absorber is vented,- therefore filled wit . orated water,' ucing differential' pressure across th tube wall. 4. The axial po shaping rod drive does not allo the APSRA to drop f owing'a reactor trip. ( s prevents possible positive reactivity ~addi n caused by the APSRA axial location.) (4 @ 0.5 ea = 2.0) REFERENCE. DAVIS-BESSE, PWR-OLC-036.01, Control Rod Drive, page 7. 001000K601 ..(KA's) ~ 1. APSR's have buttons to hold segment arms 2. APSR's have no ball checks 3. Bearing'surfac'e thicker (segments) 4. Coupling size different 5. No springs / snubber on APSR

6. A ison lengths are different

/ (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)

2. PLANT'DESIGd. INCLUDING' SAFETY AND EMERGENCY Paca 37 SYSTEtiS 1. Clecn Vaste Receiver Tanks (CVRT) 2. Primary Vater or Primary Vater Storage Tank (PVST) 3. Demin Vater or Demin Vater Storage Tank (DUST) 4. Boric Acid Addition Tanks (BAAT) 5. Concentrate Storage Tank (CST) ANSWER 2.09 (1.00) 6. . Refueling Canal g g j # er-7. Transfer Pit 1. Containment Spray pum p. Reactor Coolant Gystem 9, spe,d (uc'. Pcc1 2. Cont &iument Spray p 1-2 3. Bore.ted Watdr orage 4. Sper t Fuel ool 5. Recire. eturn from HPI' pump 1-1 returm from HPI pump 1 4 6. Rec c. (any 4 @ 0.25 ea = 1.0) REFERENCE DAVIS-BESSE, P & ID M-033, Decay Heat Removal System and Emergency Core Cooling Systems 006000K601 ..(KA's) ANSWER 2.10 (1.00') I G Cooling is provided by the amount of flow'out the pipe break t h us the RCC :nd the reeircaintien "elu-r -lere during M ggyba^ % (-h0-) @ % 6 L.m,E uado- 0% cy,4rs% cuar sos) @ Prev 4 Levb w m *O I * $ S "3 b T G t e 0,s RAeL e f.c') REFERENCE DAVIS-BESSE, TIM Vol 6, High Pressure Injection System, page 13. SP 1104.07.19, page 4. {LER 85-013} 006000K605 ..(KA's) ANSWER 2.11 (1.00) 1 i D \\ l l I l 1 (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) { l l a_--_ _ _ - 1

2, PLANT DESIGN INCLUDING SAEEIY AND EMERGENCY Pase 38 l'Of EXETEMS. .4 REFERENCE' -1 L DAVIS-BESSE, PWR-OLC-044,- Makeup and Purification, TP-1 006000108 '006000105 ..(KA's) i - ANSWER 2.12 (1.00) 4 l 1 1. Prevents thermal shock to spray nozzles. (0.50) 2. Equalizes boron concentration between the RCS and PZR. (0.50) - REFERENCE DAVIS-BESSE, PWR-OLC-32.00, Pressurizer, page 6 1 010000K401 ..(KA's) ' ANSWER 2.13 .(1.00) ESSEN Line 1 1. HPI pump 1-1 bearing housing cooling 2. DHR pump 1-1 bearing housing cooling 3. Containment gas analyzer heat exchanger 4. Emergency diesel 1-1 jacket cooling water heat exchanger 5. Decay heat removal cooler 1-1 (any 4 @ 0.25 ea = 1.0) REFERENCE DAVIS-BESSE, PWR-OLC-020.01, Component Cooling Water, page 3 and 4 008000K102 ..(KA's) l l (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)

2. PL&HT DESIGN INCLUDING SAFETY AND EMERGENCY Pcga 39 E@ Title Il' ( L 4 I ANSWER 2.14 (2.00) A 1. Provide t-3gative pressure (1/4 to 1-1/2 water) following a LOCA to: a. Annulus b. Mechanical penetation rooms 1-4 c. ECCS rooms d. Makeup pump room e. Decay heat cooler room (any three required for full credit) (0.5) A 2. Reduce airborne fission product leakage to environment by a filtration process. (0.5) B 1. Containment Purge System B 2. Radwaste Ventilation System B 3. Fuel Handling Ventilation System (any 2 @ 0.5 ea 1.0) REFERENCE DAVIS-BESSE, PWR-OLC-048.01, Containment Ventilation, pages 17 and 18. 029000K103 ..(KA's) ANSWER 2.15 (1.00) 1. If makeup to the RCS is required at a pressure higher than the HPI discharge pressure. 2. In the event of a small RCS leak, the RCS pressure is greater than the discharge pressure of the DH pumps such that HPI is required and the BWST is nearing its low level setpoint (2 @ 0.5 ea = 1.0) REFERENCE DAVIS-BESSE, SP 1104.04.24, page 28. 006000K406 ..(KA's) (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)

a 3 ,2. -PLANT-DESIGN INCLUDING ~ SAFETY AND EMERGEECX Pega 40-I SYSTEMS l-(l.%g -ANSWER 2.16 ) 1. Bus A 2. Bus % A 3. E-3 4. F-13 5. D-1 A V r dekeked S. '" ' P

v. r /- -

7. C-1 REFERENCE DAVIS-BESSE, TIM Vol 1,-Reactor Coolant Pumps TIM Vol 6, High Pressure Injection System TIM Vol 6, Decay Heat Removal System TIM Vol 7, Makeup and Purification System TIM Vol 8, Plant Air System TIM Vol 9, Component Cooling Water System 062000K210 ..(KA's) i l l l l l i (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) l

l 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Page 41 SYSTEME ANSWER 2.17 (2.00) A. 1. Wet sprinkler 2. Pre-action sprinkler l l 3. Deluge 4. Deluge-water curtain 5. Subsurface foam injection (4 @ 0.25 ea = 1.25) B. 1. Fire water storage tank [ electric fire pump] 2. Intake structure [ diesel fire pump] 3. Fire Department pumper connection (3 @ 0.25 ea = 0.75) REFERENCE DAVIS-BESSE, PWR-OLC-003.01, Fire Protectin/ Detection, pages 4, 15 -24. 086000K302 ..(KA's) ANSWER 2.18 (1.00) 1. using the makeup tank and makeup pump 2. using a boric acid pump and a primary water transfer pump 3. using an HPI pump and the BWST (any 2 @ 0.5 ea = 1.0) REFERENCE DAVIS-BESSE, SYS ORQ 0020BJ 1 and SP 1104.1. [34] 006000K602 ..(KA's) (***** END OF CATEGORY 2 *****)

ll ' 3. INSTRUMENTS AND CONTROLS-Paco 42 ANSWER. 3.01 (2.00) 1. Turbine EHC Control NOT in ICS auto 2. ' Reactor. trip 3. BOTH turbine generator output breakers open (ie. load rejection) li=it - Anodor Crowle*m'sk 4. C ::: Vg Q,*[ &< S.gQ Cross bdN3 Turbine trip 5. Fu d A cro n-l M (4 e 0.5 ea = 2.0) l c. l REFERENCE DAVIS-BESSE Examination Question Bank, Question # 03-6 ANSWER 3.02 (2.00) l 1. Lift oil. pressure 2.~ CCW flow 3. Reservoir oil level 4. Seal injection flow 5. Reactor power Core Lif t Criteria e. Tc. > SME 6. 7. Seal return valve open. B. Volt %c. > 5dPd (- 7s% 't 13.tkv) (5 @ 0.4 = 2.0) REFERENCE l DAVIS-BESSE, PWR-OLC-033, p. 19 003000K406 003000K403 003000K402 ..(KA's) (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) s

3. INSTRUMENTS ~AND CONTROLS Pago 43-4 ', ~ ANSWER 3.03 (2.00)L [0.3) for. runback [0.1) for rate. [0.1]:for limit 1. Demerator Level 50%/ min 55%- 2. ' Loss of RCP. .50%/ min 75/45% 3. Loss of-FWP 50%/ min '55% .4. FWP.Hi Dis.. Press ~.a o % to %. 3R/ min 56% (2.0)~ REFERENCE

• • • • CAF**** FACILITY DID NOT SUPPLY LESSON PLANS OR TRAINING TEXT!!

DAVIS-BESSE Examination Question Bank, Question #06-43 ANSWER 3.04 .(1.50) 1. BTU-limit on either OTSG 2. both OTSGs on level limit 3. .both feedwater loop demands in manual (3 @ 0.5 ea = 1.5) REFERENCE DAVIS-BESSE Examination Question Bank, Question #03-12 059000K107 ... ( KA ' s ) ANSWER 3.05 (2.00) 1. Overspeed 2. DG differential relay action 3. .Overcurrent 4. Manual Emergency Shutdown (4 @ 0.5 ea : 1.0) 4 i (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) J

T 3. INSIBUMENTS AND CONTROLS Pcga'44 REFERENCE 1 J DAVIS-BESSE, PWR-OLC-055.01, p. 4 Objectives, PWR-OLC-055.01, #04 064000K401- ..(KA's) ANSWER-3.06 (1.50) 1. Steam-feedwater differential pressure. .2. Low level in either OTSG 3. Loss of all RCPs 4. Low pressure in either OTSG 5. High Level in either OTSG. (5 @ 0.3 ea = 1.5) REFERENCE-DAVIS-BESSE, PWR-OLC-030.01, p. 6 Objectives, PWR-OLC-030.01, #02 & 03 061000K402 ..(KA's) op wro c.c. der (iw.b) endec,b Ao d kd Co'd Me $ 3.07 1.00 ANSWER Depress and Hor.n +% og push-buttons.for RN 1&ne b big.5] (1.0) REFERENCE DAVIS-BESSE, PWR-OLC-029.02, p. 12 Objectives, PWR-OLC-029.02, #06 061000K414 ..(KA's) ANSWER 3.08 (1.00) l Plant Vent isolation valves shut (WG 1819 & 1820). (1.0) REFERENCE ) DAVIS-BESSE, PWR-OLC-004.01, p. 17 Objectives, PWR-OLC-004.01, #05 q 071000K404 ..(KA's) l-(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) I i

,-3. INSTRUMENTS AND CONTROLS Peca 45 i I ANSWER' '3.09 (1.00) l .A. (1.0) REFERENCE DAVIS-BESSE, PWR-OLC-036.01, pages 28 & 29 L Objectives, PWR-OLC-036.01, #04 001000K401 ..(KA's) ' ANSWER 3.10 (2.00) 1. FEEDWATER CROSS LIMIT (0.25) If measured feedwater flow deviates from feedwater demand by >/= -5% (0.25), reactor demand will be modified by 1% in excess of 5% (0.25). Reactor demand will only be decreased by the actions of this limiter (0.25). 2. REACTOR CROSS LIMIT (0.25) If measured reactor power deviates from reactor demand by Yg% to + 10% (deadband) (0.25), then feedwater demand will be Sc3Y _ed by LLo omuuut of crror in exc::: ef deadband (0.25). Feedwater demand 4 maybeincreasedordecreased'tomatejgreactorpower(0.25). l'/. (e.- q l'/. M 4ke.,.sp, ) J/ 7, kr q l*4 above. b hot, 2 on REFERENCE

• • • • CAF**** FACILITY DID NOT SUPPLY LESSON PLANS OR TRAINING TEXT!!

059000K402 ..(KA's) i ANSWER 3.11 (1.00) B (1.0) REFERENCE DAVIS-BESSE, PWR-OLC-044.01, pages 4-10 Objectives, PWR-OLC-044.01, #04 004000K110 ..(KA's) (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

3. INSTRUMENTS ~AND'CONTROLE Paca 46 I ANSWER 3.12' -(1.00) l-B (1.0) REFERENCE DAVIS-BESSE, PWR-OLC-039.02, p. 39 Objectives, PWR-OLC-039.02, #10 012000K406 ..(KA's) ANSWER 3.13 (2.00). A. 50 psig B. 145 psig C. 50 psig D. O psig (4 @ 0.5 ea = 2.0)- REFERENCE DAVIS-BESSE, PWR-OLC-027.01, pages 13 & 14 Objectives, PWR-OLC-027.01, #06 041020K418 041020K417 ..(KA's) ANSWER-3.14 (1.50) BTU limits significantly reduced to both FW loops resulting in FW-reduction to both OTSG's (0.5). The indicated Tave is low, rods ~ pull to increase neutron poner (0,5). the reactor then trips on high RCS pressure (0,5). REFERENCE

• • • • CAF**** FACILITY DID NOT SUPPLY LESSON PLANS OR TRAINING TEXT!!

016000K403 ..(KA's) 1 1 (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) l k

3. INSTRUMENTS ~AND CONTROLE-Pass 47 d' . ANSWER 3.15 (1,50) The' basic pump control is taken as a feed-forward control. signal (0.5) and a FW valve delta P feedback signal. The total flow ' demand signal is modified by addition of the delta P feedback signal (0.5) so that large or rapid load changes will' change the pump speed accordingly (0.5). (1.5) REFERENCE -

• • • • CAF**** FACILITY DID NOT SUPPLY. LESSON PLANS OR TRAINING TEXT!!

STM 32 " Integrated Control System" pg 32-84. 059000K405 059000K406 ..(KA's) ANSWER 3.16 (1.00) 1. Absolute (0.25), is sensed by reed switches (0.25). 2. Relative (0.25), is sensed by a stepping motor in parallel with the CRDM motor phases (0.25). (1.0) REFERENCE-DAVIS-BESSE, PWR-OLC-036.01, pages 28-32 . Objectives, PWR-OLC-036.01, #04 014000SG07 ..(KA's) J 1 ANSWER 3.17 (2.50) 1 A. 1. Normally closed supply breaker opens. 2. EDG starts but does not tie to bus. 3. The other TWO supply breakers get a " prevent close" signal. 4. All load breakers trip. (4 @ 0.5 ea = 2.0) B. The EDG output breaker closes on the bus. (0,5) I (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) t

,3. INSTRUMENTS AND CONTROLS Pace'48 4. REFERENCE DAVIS-BESSE,.PWR-OLC-054'.01, pages 8-9 Objectives, PWR-OLC-054.01, #04 062000K403 062000K402 ..(KA's)- 4 l f l kASTR CPY (***** END OF CATEGORY 3 *****)

] 4. PROCEDUBES'- NORMAL. ABNORMAL. EMERGENCY Paga 49-AND RADIOLOGICAL CONTBQL ANSWERE 4.01 -(1.00) 1. Reactor trip -2. SFAS trip (W/E level 1 or CTMT RAD) '3. SFRCS trip L4. 'OTSG tube rupture 5. -When directed by another procedure .6. Operator judgement (any 4 @ 0.25 ea = 1.0) REFERENCE-DAVIS-BESSE, GOP-OLC-003, Attachment 2.02, page 7 and lesson objective # 01. 194001A116 ..(KA's) ANSWER 4.02 (1.00) 1.. Work has been completed 2. All personnel holding clearance have released clearance 3. The SS/ Assistant SS have determined the equipment is ready for service. (3 @ 0.33 ea = 1.0) REFERENCE DAVIS-BESSE, ADH-SRO-005.01, Safety Tagging Procedure I I AD-1803.00 194001K102 ..(KA's) 1 ( i l l l (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

L,

41 PROCEDURES'- NORMAL.' ABNORMAL.' EMERGENCY Pago 50 AND RADIOLOGICAL CONIRQL-e Sa

• Us(1.00)'d %3 m f.Il 4 hase..

4.03 ' ANSWER-t of 14 ' Operating the Nuclear Steam Supply System, turbine gene

tor, r auxiliaries, and all othe equiprn.;nt to maintai cont ous production with maximum safety and effi ency.

2. Reporting t he Assist. SS information import, t to the safe and efficient ration of the station. / ,3. Being alert and atte ive to the instru ntation and controls in the' control room an ontrol room ea. 4. Maintaining control of personn entering the control room, 5. Maintaining the Reactor Ope tor's control room reading-sheets, and the Primary ant Status Bo d. 6. Shutting down the actor when he determines at the safety of the reactor in jeopardy or when operating parameters ceed..any reactor protection setpoints nd automat shutdown does not occur. 7. ntaining as NRC Reactor Operator's licence on the unit. (any 4 @ 0.25 ea = 1.0)- REFERENCE. Ru>lk, DAVIS-BESSE, AD 1L&32.00, Eif t Opr:tien(page 9. lt39.co cAdsk epug},Q 002000SG01 ..(KA's) 1 (e **** CATEGORY 4 CONTINUED ON NEXT PAGE *****) ) l l I ..z____._ __________________________,J

i[ , I (1 ads.8 E SSE lBNIZi@9E@tG QH5X5G@ mig PN 20@ REiriSiON CCE0t,PE %.,Nia l Q f Sh * - ' Station Opera 1jons ..i; } 9 of 84 -16~ 'AD 1839.00 a s z y

• :5 8.)s

-\\ Observing recorders and indicators within th'e control. room i for safe operation within prescribed limits; performing necessary switching, regulating and synchronizing optrations; J. maintaining system frequency, as scheduled; .} S.8.2 Checking the operations of equipment; making adjustments of equipment to ensure efficient operation, initiating switching i sequences required by equipment. loading and clearing equip ' L 1, ment in trouble; 5.8.3 Performing various pre-startup, surveillance, and. stand-by i' j operation, checks on instruments, alarms and equipment; l 5.8.4 ' Assisting in the pre-critical check-off and making opera-f

• tional adjustments as. required; l

5.8.5 Diagnosing and correcting abnormal opera-ing conditions; (

• directing and instructing other personnel in procedures

} ) necessary to correct abnormat operating conditions or. prevent .their recurrence; i i 5.8.6 Responsible for shutting the reactor down when he determines that the safety of the reactor is in jeopardy or when operat- ) ing parameters exceed any reactor protective set points and l automatic shutdown does not occur; 1 5.8.7 Keeping working area and equipment in a neat, clean'and [' ! presentable condition; j 5.8.8 l Performing and/or assisting.in the performance of routine and~ special' testing of station equipment and systems; 5.8.9 Maintaining required' logs and calculations, assisting in the l preparing reports, computing station data, drawing curves and y charts, preparing and reviewing, reporting errors in, and l making suggestions for improvements of the relevant Station 'j Procedures and Tests; 5.8.10 Instructing and training employees in the duties of Reactor Ope rato r; 5.8.11 Performing the duties of Equipment Operator; t J 5.8.12, Performing duties required in Station Emergency Procedures; 5.8.12 Working with employees of other classification and performing other duties in accordance with the section of the Foreword k to the Job Manual entitled " Duties"; and l H l- } t l 5.8.14 Maintaining a Reactor Operators License in accordance with the Davis Besse qualifications program. t-Procedure Text i Page 9 of 79 l I l

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a

4.

PRQQEDURES : ' NORMALi.... ABNORMAL. EMERGENCY Peca 51 L'

AND RADIOLOGICAL CONIBQL j,. ( t . ANSWER. 4'.04 (1 50)

1. -

Load Changes.(greater than 10 MWe) l l

2.. Reactivity changes 3.

Ocurrence.of alarms pertaining to reactor core conditions 4. All releases of radioactive wastes 5. -Equipment status changes E 6. Performance of significant procedural steps 7. Suspected reportable occurrences 8. Execution of the Emergency Plan 9. ' Performance of surveillance testing (any 6 @ 0.25 ea.= 1.5) REFERENCE DAVIS-BESSE, AD 1839.00, Station Operations, page 24 000001SG01 ..- ( KA ' s ) ANSWER 4.05 (2.00) 1. a = 16 d = 24 b = 16 e = 72 g c = 24 f=8 r. (6 @ 0.25 ea = 1.5) '2. Davis-Besse Plant Manager or his designee of higher position (0.5) a l a (***** CATEGORY 4 00NTINUED ON NEXT PAGE *****) i i I

_(. PROCEDURES - NORMAL.' ABNORMAL. EMERGENCY Pega 52 -AND~ RADIOLOGICAL CONTROL m REFERENCE-DAVIS-BESSE, AD 1839.00, Station Operations, page 54 000001SG01 ..(KA's) 1 ANSWER 4.06 (1'.50) 1. 1250 mrem 2. 3000 mreta -3.- 4500 mrem (3 @ 0.5 ea = 1.5) ' REFERENCE DAVIS-BESSE, HP 1601.01.10, page 2. 194001K104 ..(KA's) ANSWER 4.07 (2.00) 1. Control rod safety group 1 is at its upper limit 2. _The makeup-and purification system is disrupted by a loss of letdown, loss of makeup pump, loss of makeup flow, etc. 3. If control rod group position indication, neutron count rate, or other reactivity indications are behaving in an erratic unexpected manner. 4. .If the time calculated for the operation varies significantly from actual operation time. (4 @ 0.5 ea = 2.0)

REFERENCE DAVIS-BESSE, SP 1103.04.8, Boron Concentration Control, page 2 004000SG01

..(KA's) (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

.- 4. EBQQEDUBES~~ NORMAL. ABNORMAL. Et1EBGENCY Pega 53 AND RADIOLOGICAL CONTROL ANSWER. 4.08 (1.50) '1. 'An operating feedwater pump (MFP, AFP, MDFP) 2. Feedwater to either STM generator 3. With STM generator level increasing to, at, or above proper setpoint. (3 @ 0.5 ea = 1.5) REFERENCE DAVIS-BESSE, EP 1202.01, page (caf) [#23] 056000SG13 ..(KA's) -ANSWER 4.09 (2.00) S Tc/p 5 Fees en hi$l fuel

4. Vu*,? P w sfR.Ls W 3

1. Start a second makeup pump 7, g g g g 4 g,3 yd V M (p 6 s J a. 2. Isolate letdown 3, % g Acu.= 1,Jers

i. bkod M*N he%w 3

Manually trip reactor se. r A s q g y. 4. Verify all control rods inserted (except aprs) (4 @ 0.5 ea = 2.0) -REFERENCE i DAVIS-BESSE, AB 1203.12, page 5 [47] 002000SG14 ..(KA's) I 1 ANSWER 4.10 (1.00) I i 1. Adequate subcooling margin has been restored 2. Pressurizer level !.s greater than 100 inches and increasing (2 @ 0.5 ea = 1.0) (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) l

4. PROCEDURES - NORMAL. ABNORMAL. EffEBGENCY Pega 54 .AND RADIOLQQICAL CONTROL 4 REFERENCE DAVIS-BESSE, EP 1202.01 [62] -006000K402 ..(KA's) ANSWER 4.11 (1.00) 1. Visually looking at the valve stem 2. Physically applying force to the handwheel'in the closed direction. (2 @ 0.5 ea = 1.0) REFERENCE DAVIS-BESSE, ADM-SRO-005, lesson obM1 [24] {LER 85-017} 000022SG07 ..(KA's) ANSWER 4.12 (2.00) A. = C (0.5) B. 1. start a second makeup pump fully open MU-32 (makeup control valve);,k F Mu311t W 2. p 3. shift makeup pump suction to the BWST vi vivolus M" 0071-(makeup pump 3-way suction valve) (3 9 0.5 ea = 1.5) REFERENCE DAVIS-BESSE, GOP OLC 003, OBJ 1 and EP 1202.01 [9] [11] 10CFR5521J ..(KA's) 1 l (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) L______z__________

c. ii PRQQEDURES - NORMAL. ABNORMAL. EMEBGEEQY Peco 55 1 AND RADIOLQQICAL COETROL ~ j i A WER 4.13 (1.00) outimfpipe etions as directed by applicable procedure 1. 'Carr 2. Performs actions ed by Assistant Shift Supervisor 3. Acknowledges in a positive er the completion of actions directed by the " Procedure Reade uring implementation ,of EP 1202.01 4. Made appropriate recommendations to Assistant Shift Supervisor. (4 @ 0. - 1.0) REFERENCE DAVIS-BESSE, GOP-OLC-003.00, PAGE 15. 194001K109 ..(KA's) ANSWER 4.14 (1.00) to.D MU-19 would be fully open energizing the bus would cause the g makeup pump to. start causing maximum flow to the RCP seals Co.53. This would thermal shock the seals 5bm[b.s]. (1.0) REFERENCE DAVIS-BESSE, GOP-OLC-003, Attachment 3.02, page 3 and lesson ojective #3 1 062000A104 ..(KA's) (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

4. PROCEDURES'- NORMAL. ABNORMAL. EMERGENCY Para 56 AND RADIOLOGICAL CONTROL-V r. 1 j:. 1 ANSWER. 4.15 (2.00) i

n.

• L L

1. . Control rod groups 1, 2, 3, and 4 must be 100% ( ) withdrawn for.any. feed and bleed operations. O. 6 (. 2. Control rod group 5 must be greater than 25% withdrawn. (976(

3. # D g or pc-determices--the nocussity of a thied

.~w........ a, If the reactor power is greater than 15% full power, the actual position of group 7 must be more than 5 percent above or below the nominal

o. G c position on the' transient rod position band

( &v24,) t ~~1f reactor power is less than 15% full power J group 7 position has no effect on the permissive. 007253' ( SIMILAR WORDING ACCEPTABLE ) 1 REFERENCE DAVIS-BESSE, SP 1103.04.7, Boron Concentration Control, page 11 004020K401 ..(KA's) d ANSRER 4.16 (1.50) l 1. 90 seconds (+/- 5 seconds) (0.75) 2. 4.5 minutes (+/- 0.5 minutes) (0.75) 1 REFERENCE l DAVIS-BESSE, SP 1103.06.13, Coolant Pump Operating Procedure, page 7 008000SG10 ..(KA's) l l I ANSWER 4.17 (1.00) 1. Less than 20% of full scale (0.5) 2. 80% of full scale . (0.5) (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

4. PRQQEQHBES - NORMAL. ABNORMAL. EMERGENCY Peca 57-AND_B&DIOLOGICAL CONTR'3L o REFERENCE. DAVIG-BESSE, HP. 1601.03.10, page 5. CFR5055211 ANSWER 4.18 (1.00) 1. 'Between 5 x 10(-8) and 5 x 10(-7) amps. ogemids Leee tb-f.1 DPM 0 2. 1 (2 @ 0.5 ea = 1.0) i REFERENCE' DAVIS-BESSE, Plat.e Startup, PP1102.02.25, page 66 010000SG07 ..(KA's) ~ NASTER CO)V (***** END OF CATEGORY 4 *****) (********** END OF EXAMINATION **********)

o c MASTER COPY U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _D8yl@-@gggg_____________ i REACTOR TYPE: _PWR-@gW1ZZ______________ DATE ADMINISTERED: _ _8 _7 _/ 0 8 _/ _1 7 _ _ _ _ _ _ _ _ _ _ _ EXAMINER: _@@LygR@2_G _____________ CANDIDATE: INSIBUGI19NW_I9_CBNQ1991gi Use separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination,,, starts.~ ,y f' i % OF CATEGORY % OF CANDIDATE'S CATEGORY __Y6LUE_

IDIGL,

___SGOBE___ _YGLUE__,______________CBIEGQBy_____________ 25,?s _dus2E__ _29.i_bd 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS,'AND THERMODYNAMICS A 2, C-299rk2__ _2Zzag ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 2YO _@4th7T_ _22196 ________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 23.?5 _Elt25__ ~22x@Q ________ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS re2:e2__ Totals Final Grade AlI wark done on thin examination is my own. I have neither given nor received aid. j ) Candidate's Signature s l I

l}. g;'

f 7.,.

..I INRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS ~ Daring the administration' of this examination the following. rules apply: 11 Cheating.on the examination !means an automatic denial of your application t and.could result'in more' severe penalties. ~

2.

Restroom trips are toLbe limited and only.one candidate at a time may. leave.. You'must avoid.all contacts with:anyone outside the examination room.to' avoid: even-the ' appearance. or possibili ty of ch' eat i n g. J 3. Us. black. ink or ' dark pencil ggly. to f acilitate. legible reproductions. 4. Print your.name in the~ blank provided on the cover sheet'of the examination, j5. Fill in the date on the cover sheet of the examination (if necessary). 6. 'Use only the paper provided for answers. 7.-(Print your name in the. upper right-hand corner of the.first page of ggch section of the answer sheet. 8. Conse.cutivcely number each answer sheet, write "End of Category.__" as appropriate, start each category on a agw page, write gely gg: gag gidg of.the paper, and write "Last Page" on the last answer sheet. 9. Number each answer as to category and number, for example, 1.4,<6.3. 10./ Skip at least th gg lines between each answer. t 11.. Separate answer sheets from pad and place finished answer sheets face down on your desk or table, j 1 l

12. Use abbreviations only if they are commonly used in facility litggetutg.

1 l

13. The point value for each question is indicated in parentheses after the l

) question and can'be used as a guide for the depth of answer required. { J L#

14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.

-15.. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK. l

16. If parts of the examination are not clear as to intent, ask questions of the gd.cmicer_ on1 y.

17. You must sign the statement on the cover sheet that indicates that the work is your &a and you have not received or been given assistance in completing the examinati on.

This must be done after the examination has been completed. [4 1 l l .O

E i

10. When you conipl ete your ex ami nati on, you shall a.

Ansemble your examination as follows: (1). Exam questions on top. (2) Exam aids - figures, tables, etc. (3) Answer pages including figures which are part of the answer. b. Turn in your copy of the examination and all pages used to answer ci estions. the examination t c. Turn in all scrap paper and the balance of the paper that you.did not use for. answering the questions. d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this.aren while the examination is still in progress, your license may be denied or revoked.

A__IBE9BX_DE_NUCLEGB_E9BEB_E60NI_0EEBSIlgN _ELUlpg3_OND PAGE 3 IUEBMQDyNSL11gs 4 QUESTION 5.01 (2.00) How would the actual critical rod position-vary from the estimated critical rod position (ECP) for EACH ot the following situations?(Higher,. Lower, or Remain The Same) Inc_lude a DR10 explanation of WHY this occurs. an ECF is calculated using zern unon 1. After a tr ip f rom 1007. par.er, reactivity for a startup 8 hourr, at ter the tr ip. ~ (0.5) 2. The actua] t>orori concentr at t an as 100 p p ro lower thall the value used for the EUP. (v. b) ~.. "J n EFT U i st uneri )n the ECF 1 rmt ea d at the actual 200 EFPD. (U.D; 4. the soitrce count rate has decreased from 20 cps to 10 cpu between the time t he ETF won calculated and the beginni tig of the clar top. (O.U) DUbbil I ON b.OP (1,00 LIST the two (2) f ac t or w (other than uncer tai nty factors) which are mu]tiplied togethe-ta calcuiate the HucIear Heat FJ ux Hot Channe) Fattor (FU). QUEB'110N b. O!: ().GO) ilet ' of Ihe 1 oper n'. t cine 1 cond)t2nos necessary to maintain the Hot Cliasin el Fxbu s within design l i mi t s as st.ated in T.S.' (RJi.;bi *+iN '. 0 4 (1.25) .i. How Ao wi io i-13'? p r ed t umo avid r van ved in the core? Gi ve decay t i nus. (1.(D 7. What per c eiit of oil tissions, eather directly or through decay, pr oduce Xenon? (0.25) (***++ CM E6Uic i 05 CONTINUED ON NEXT PAGE

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i I i t 5. THEOHY OF NUCLEAR POWER PLANT OPERATION FLUIDS _AND PAGE 2 2 _T H_ E_ R_ M _O D _Y _N A M _I C_ S y QUESTION 5.95 (1.25) 1. LIST three (3) components which make up the power defect. (0.75) 2. DESCRIBE hov Power Defect changes over core life? (0.5) QUESTION 5.06 (1.5^) LIST three (3) reasons for establishing the regulating group insertion limits per T.S. bases. QUESTION 5.07 (2.00) 1. Define SHUTDOWN MARGIN Per T.S. (1.0) 2. List 4 of the 5 parameters for calculating SHUTDOWN MARGIN as listed in SP 1103.15 worksheet 4. (1.0) OUESTION 5.08 (2.25) 1. Define Beta core. (0.5) 2. LIST the following isotopes U235, U238, and Pu239 in descending order of the value of their beta. (0.75) 3. Explain how and why beta core changes over core life. (1.0) l ) ( (***** CATEGORY 05 CONTINUED ON NEXT PAGE

• ++**)

1 l L.

'l 1 ) ) Sc__IUE0BY_9E_NyGLE86_EgWEB_E68N1_QEEB8IlON _ELy1D@3_@ND PAGE tl 2 IMESdODYN@dlGS QUESTION 5.09 (1.00) l The reactor is shut down by 6% delta k/k with a source neutron count rate indication of 50 cps. Rods are withdrawn to raise the source range indication to 285 cps. CALCULATE the value of reactivity (+ or -) when counts are 285 cps. (SHOW all work) OUESTION 5.10 (2.00) i Db.W 177-fuel assembly plants have 52 SPND detector strings. The detector strings are inserted in the central instrument tube of the selected fuel assemblies. 1. List the 3 (three) major components and their quantity in a single SPNP detector string. (1.5) 2. EXPLAIN how a SPND detects and indicates a neutron flux? (0.5) i i OUESTION 5.11 (1.00) The pressurizer PORV is leaking by during operation at 85% power. Assuming a Quench Tank pressure of 20 psia and saturation conditions i in the pressurizer corr esponding to 2240 psia, what is the quality of steam on the downstream side of the PORV? Show all calculations. J l QUESTION 5.12 (1.50) LIST 3 of the 4 methods used to verify that natural circulation heat i removal to the OTSG's is present according to PP 1102.10 Plant l Shutdown and Cooldown (Include limits as appropriate) ) (1.5) I a 1 1 i (**+** CATEGORY 05 CONTINUED ON NEXT PAGE

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J 'Ut__ISEgBY QE_UUC6588_EQWEB_E68NI_QEEB811gN _ELUlpg2_QU2 PAGE t 1 ISEBdQDYN6dicg l.0 QUESTION 5.13 W 1. If the' temperature of the coolant entering the suction of a ))b[h centrifugal pump is increased, State how the NET POSITIVE SUCTION HEAD REQUIREMENTS would respond. (Choose one) (Increase,' Decrease,'or Remain The Same.) (0.5) 2. Given that if the speed of a centrifugal pump increases, the available Net Positive Suction Head decreases. WHY7 l (1.0) QUESTION 5.14 (1.00) A cool.ing water pump operates at 1200 rpm, has a capacity of 800 gpm, and requires 60 kw of power during normal operation. Calculate the power requirement, if the rpm was increased to 1800. (1.0) OUESTION 5.15 (.75) LIST the three (3) DNB parameters which must be maintained within limits in T.S. 3/4.2.5, "DNB PARAMETERS." (0.75) DUESTION 5.16 (1.50) Given the f ollowing data: ( i core output 2700 MW f Th 606 deg. F l Tc 554 deg. F 1 Tave 580 deg. F l superheat 35 deg. F ] OTSG Header pressure 900W Temperature of the feedwater 480 deg. F OTSG Lavels 88% q I 1. Calculate the primary coolant flow. i i i (**+** CATEGORY 05 CONTINUED ON NEXT PAGE

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7 '5 __ISEQBY_9E_UWGLEBB_EQWEB_E66NI_gEEB611gN _E6WlQ@i_6dQ PAGE tl t t ISESdQD198dlG9 QUESTION 5.17 (2.25) 1. E>c p l ai n what is occurring in the following heat transfer regions 1. Nucleate boiling region (1.5) 2. Film boiling region 3. Superheat region 2. As power in increased from 0% to 100% power, state whether the area-of the. heat transfer region. (increasen,. decreases, or stays the same.) 1. Nucleate boiling region (0.75) 2. Film boiling region 3. seperheat region DUESTION 5.$8 (1.00) List thee four (4) thermodynamic conditions needed for natural circulation to occur. ) (***** END OF CATEGORY 05

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!s __ELONI..SYSIEdg_DESlgg3_ggNIBQ62_@Np_INSIBgMgNISIlgN PAGE g I 1 3 1 j I i QUESTION 6.01 (1.00) ') i Match the following SFRCS equipment to their power supply. j l a. SFRCS CHANNEL 1 a. YAU (1.0) b. Y1 l i b. SFRCS CHANNEL 4 c. D.1 N e. D1P c. SGLIC 2 f. YBU ]j g. Y2 d. SGLIC 3 h. D2N j 1. D2P (4 @ 0.25) QUESTION 6.02 (1.25) LIST the (five) 5 parameters that will cause a SFRCS FULL' TRIP, including setpoints if applicable. QUESTION. 6.03 (2.00) When the SFRCS is actuated, OTSG 1evel control is dependent upon a three position mode switch located on the operator's control board of which the ICS control position is blocked from operation. LIST the other (two) 2 positions and EXPLAIN how they are used at Davis Besse to control. level. (include level set points) (2.0) l QUESTION 6.04 (1.00) There are (two) 2 identical CTMT Hydrogen Analyzers, each capable of sampling one of four different points. STATE where these sample points are physically located in containment. (1.0) QUESTION 6.05 (1.00) L Describe the flow path from the Containment to the Hydrogen l Recombiners back to the Containment. Identify all major components and L inter connecting systems. L. (***** CATEGORY 06 CONTINUED ON NEXT PAGE +++++)

ef__EL8NI_SYSIEdg_DEgl@N3_ CON 1B962_BND_INSIBUdENI@IlgN PAGE t' ' 4 4 OUESTION 6.06 (2.00) 1. LIST the three (3) inputs to ARTS.and.from where the signal is being sensed, other than SFRCS. (1.5) 2. WHICH one of the SFRCS at:tuation signals will not cause an ARTS trip? H (0.5) OUESTION 6.07 (1.00) j 1. TRUE or FALSE. It takes a full SFRCS trip to cause an ARTS trip. (0.25)

2. When Channel f[of ARTS trips, what does the opening of the output contacts KA and KB do, and what component does it cause to operate?

(0.75) QUESTION 6.08 (1.50) The MSR' drain tank has a high level turbine trip associated with it. This trip is 3 inches below the bottom of the MSR. 1. State the purpose of this trip. (0.5) i 2. Describe how the steam / water gets from the MSR entry point to the-MSR drain tank. (0.5) 3. State where the normal level control valve for the MSR drain tank drains to. (0.25) 4. State where the high level dump valve for the MSR drain tank dumps. (0.25) J (+**** CATEGORY 06 CONTINUED ON NEXT PAGE + ++)

6 __E60NI_@y@lgdS_QEgigN _ CON 16063_GNQ_lNSIBudENI@llgN PAGE E 5 3 k QUESTION 6.09 (1.25) 1. ON what component in the reactor vessel internals are the INTERNAL VENT VALVES located. (0.25) 2. What ss the purpose of the Internal Vent Valves. (0.5) s 3. At what pressure do the Internal Vent Valves begin to open. (0.25) 4. At what pressure are they fully open. (0.25) OUESTION 6.10 (1.50) 4 1 I LIST fi ve (5) of the eight (8) Trips that will cause an automatic shutdown of the Control Room Ventilation System.(Set points not required) (1.5) I l DUESTION 6.11 (3.00) STATE the setpoint and the reason / basis f or each of the following RPS Tripe. (2.0) ) 1 1. High RC Bldg Press (Containment Press. High) 2. RC High Temperature 3. RC High Press DUESTION 6.12 (1.00) LIGT the four (4) trips which are removed when the RPS is placed into " Shutdown Bypass" mode. (1.0) (***** CATEGORY 06 CONTINUED ON NEXT PAGE

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7,---. 1 sg__E60NI_EYSIE d@ _ Q E@l@N1_C QWlBOL2_6N Q _1NgIBydEUI61,1,QN PAGE if f QUESTION 6.13 (2.00) DRAW the control rod withdrawal inhibit and control rod withdrawal inhibit bypass logic of.the nuclear instrumentation (2.0) QUESTION - 6.14 ' (1.00) 1 1. What one system is the major entry point for oxygen being introduced into the Gaseous Radioactive Waste Syst'em. (0,5) j 2. FILL IN THE BLANK If radiation monitor sample flow drops below 6. + 1.2 SLPM, an alarm will actuate except when HV 1819 and HV 1820 are (0.5) \\ QUESTION 6.15 (1.00) During nor mal power operation the NNI System supplies a Unit Tave signal to the ICS. DEBCRIBE What will cause the Auto / Manual Transfer Switch to automatically transfer to a. loop Tave. . ( 1. 0 ) ...x, QUESTION 6.16 (1.00) Which NNI c ab i nt't contains all of the NNI i nstruinent at i on which is not redundant? DUESTION 6.17 (1.50) What provides the pri mary closi ng f orce between the rotating and the st ati onarv seal faces of the RCPls for the f ollowing conditions. 1. System start up at Low Pressure (0.75) 2. Normal operating pressure. (0.75) s s (***** CATEGORY Oh CONTINUED ON,t;SXT PAGE *****) s

l h 6 __eLeNI_@y@I@d@_Qg@l@N2_CONIBg61_6UQ_IN@lOgdgNI@IlgN PAGE 1? 7 l M{O OUESTION 6.18 ) l' 1. What determines control bleed off flow under normal conditions? (0.5) 2. WHAT recent modification was made to increase the amount of fluid between the RCP seals to reduce seal failure. (0.5) f I. What one specific 3 failure mechanism did this modification address? (0.5) l QUESTION 6.19 (.50) i TRUE or FALSE In auto, the first stage pressure feedback will drive (reset) out if the turbine is transferred to ICS control or if it goes below 20% of l first stage precuure, j (0.5) ggjh-f QUESTION 6.20 (.50) 10A Presently only one turbine back is in service. STATE the back. (0.5) DUESTION 6.21 (2.00) When the Component Cooling Surge Tank Level decreases to 35 inches, level switches will close or block the opening of four (4) valves or group of valves in the Component Cooling Water System. LIST four (4) of the six (6) major components that lose CCW flow upon low CCW Surge Tank level. (+**++ END OF CATEGORY 06 +++++)

.I J . Zz__EB99EDWBES_;_bpBd863_8DNQBd862_EMEBgEdgy_8ND PAGE 1; BSD196991986_99NIBOL i l5 OUESTION 7.01 i EP.1202.01, LACK OF ADEQUATE SUBCOOLING MARGIN has you. secure RCP'S within two (2). minutes of loss of subcooling margin. What are the BASES f or that action include the bases for the two (2) minutes? OUESTION 7.02 (1.50) STATE the failure mode of the Make Up Flow Control Valve MU 32 for the following conditions. (f ail open, half open, close, as is) a. Ins,trument air pressure 65 psig. (0.5) b. 1.A. > 65 psig and loss of NNI X AC. (0.5) c. I. A'. > 65 psig and loss of NNI X DC. (0.5) QUESTION 7.03 (2.00) 1. STATE the limits given in AB 1203.40, Steam Generator Tube Leak for _OTSG Tube to Shell delta T. for the following: 1. Normal tensile tube-to-shell delta T (tubes colder) (0.5) 2. Compressive tube-to-shell Delta T (shell colder) (0.5) 2. State how to determine (include instrumentation): 1. Tensi l e tube-t o-chel l Delta 1 (0.5) 2. Compressive tubo-to-shell Delta T (0.5) DUESTION 7.04 (.50) TRUE or FALSE During normal operation, a delta T of up to 50 deg. F is expected from l the outer peripheral incore T/Cs to the center of the core incore L T/Cs. The shape of the temperature profile will parallel the prof 21e of the neutron flux across the core during normal operation. (0,5) i 1 (***** CATliGORY 07 CONTINLIED ON NEXT PAGE ** * **) ( L______:_________.

) e t .Zt__BBgGEDUBE@,;_UQBd863_6@NQBdG6,_EDEB@Edg2_6UQ PAGE 1~ 80 Dig 6Q@lG86_GQUIBQL QUESTION 7.05 (2.00) EP 1202.01 SPECIFIC RULE 2.3.1 STATES: HPI may be stopped if the LPI system has been started and flow has been > 1000 gpm/line for > 20 minutes. 1. What is the basis for 20 minutes? (0.75) i 2. What is the basis for 1000 gpm? (0.75) 3. What is the basis for each line? (0.5) OUESTION 7.06-(1.00) Per SPECIFIC RULES 4.1 of EP 1202.01.04 During a LOCA, instrument compensation for elevated containment temperatures may be necessary. Hcn9 is the containment air temperature determined? OUESTION' 7.07 (1.00) What instrument air pressure are you required to trip the reactor and actuate SFRCS, per AB 1203.36 LOSS OF INSTRUMENT AIR 7 OUESTION 7.00 (1.00) In AD 1203.24, Circulating Water Pump Trip / Rupture, is the f ollowing caution: "Do not immediately start the comp) i mwntary circul ating water pump if the running pump was on cooling tower bypass. Condenser damage may occur." EXPLAIN the reason #or the " CAUTION" l 1 l ) OUESTION-7.09 (1.50) k,'/ g /[C/ / In AB 1203.40, 1 TEAM GENERATOR TUBE LEAK, if the SGTL is greater than 50 gpm the actions are different then for a leak less than 50 gpm. ( WHAl is the major difference? Include the setpoint at which these different actions occur. 1 I (***+* CATEGORY 07 CONTINUED ON NEXT PAGE +++++-) L_ __ _ -.

r a Zr__EBgggpuBgg_ _BgBd862_8@N9Bd662_EMESggBgy_@NQ PAGE if RADIOLOGICAL CONTROL i 1 1 i QUESTION' 7.10 (1.50) l EP 1202.01, Emergency Procedure for OTSG Tube Rupture, has you compare RE600 and RE609 readings to determine the leaking OTSG. The Main ' Steam Line Rad Manitors will not respond in the analyze mode with the Reactor Shutdown. Why won't they respond and what must be done to restore the monitors to service? 'I OUESTION 7.11 (2.00) On-the curves for RC PRESSURE / TEMPERATURE LIMITS there are two (2) curves for Fuel Pin In Compression: one for forced flow and one for natural circulation. 1. WHAT are the Fuel Pin Compression Curves intended to prevent (1.0) 2. WHY are the curves shifted ie. natural circulation to the left of forced flow. (1.0) QUESTION 7.12 (3.00) AB 1203.19, PRESSURIZER SYSTEM ABNORMAL OPERATION, lists three (3) methods of indication of PORV position and status. I LIST the three (3) methods and the information provided or derived by each. I DUESTION 7.13 (1.50) l 1. LIST the two (2) entry conditions per the Emergency Procedure EP 1202.01 other than RPS, SFAS, SFRCS Trip, or SG Tube Rupture. l (1.0) l 2. What is the exception to the SFAS Trip requiring entry into this procedure? (0.5) (**+** CATEGORY 07 CONTINUED ON NEXT PAGE

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-PAGE 1El S69196991G66_G9NIS96 QUESTION 7.14 (2.00) Answer the'following questions per PPiiO3.08, Approach to Criticality; 1. What is the maximum difference between actual and desired baron concentration prior to going critical? (0.5) 1 2. RCS temperature must be varified >525 F Within HOW MANY minutes of goi ng critical ? (maximum amount of time). (0,5) 3. Within HOW MANY minutes prior to withdrawal of any regulating rod, during an approach to criticality, shall each safety rod. be determined to be fully withdrawn? (0.5) 4. The maximum startup rate with no rods being pulled is one decade per minute (dpm), if rods are being pulled the rate l limit is increased to 1.5 dpm. WHY the differencein the allowable startup rate? (0,5) l DUESTION 7.15 (2.00) I l i TRUE or FALSE i Per AB 1203.32 " Steam Generator Feedwater Chemistry Out of Spec" answer the f oll owing questions true or false. 1. If the cation conductivity is greater than 2.0 micro-mho/cm(at the i SG Inlet) unit shutdown is required immediately. 2. The most likely cause of a feedwater conductivity problem would be a condenser tube leak. 3. Alarm "CNDS DEMIN DUT CONDUCT HI" would normall y be an indication of resin depletion of the condensate polishers. 4. For tube / tube sheet leakage in the condenser to determine which part of the condenser is leaking, Chemistry personnel should sample j from the grab sample locations. l l l (*+*** END OF CATEGORY 07

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I l 1 ' B __OQt!ldl@lBOIli>E_ EBgCEQUBES._CQUQlIlgNSz_BNQ_Ll(11IOllgN@ PAGE 14) g 3 t DUESTION 8.01 (1.00) T.S. 3.11.2.5 EXPLOSIVE GAS MIXTURE 1. WHAT two (2) gases are they concerned with? DUESTION 8.02 (2.SO) List five (5) conditions which oculd result in declaring a movable control assembly inoperative in accordance with Tech Specs. Do not include possible causes for the condition in your answer. l UESTIGN 8.03 .'cg G Per AD 1850.04, Post Accident Radiological Sampling and Anal ysi s, Which of the following in the responsibility of the Emergency Plant Manager? (Choose one) Responsibility es a. Cont act ing the Emergency Control officer at LRC. b. Authorizing that a PASS sample be taken. c. Coordination of sample movement to and from the analysis facility. d. Determination of maximum allowable personnel dose to obtain the sample. DUESTION G.04 (1.50) Per AD 1839.04, Shift Technical Advisor, During what three (3) p1anned evolutions is the STA required to be in the Control Room? DUESTION O.OD (1.50) 1. If a Safety Limit is violated, what 2 notifications are required au } given in AD 1839.00, Station Operations, after completion of T.S. required ac t i ont'> f 2. Who author ines resumption of operation? (**++* CATEGORY 08 CONTINUED ON NEX7 PAGE

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L 0 __8Dd1NISIBOIlyE_EBQCgpgBES _QQNpillggg3_8ND_Lidll61196@ PAGE 10 4 2 QUESTION 8.06 (1.50) i LIST the documents relating to Surveillance Testing the SS and 1 Assistant SS review prior to turnover. l QUESTION 8.07 (1.50) According to AD 1803.00, Safety Tagging, restoration to service shall -not begin until what three actions are' performed? QUESTION 8.08 (1.50) According to AD 1823.00; Jumper and Lifted Wire Control. '1. Install ation of a jumper on Crit 2 cal systems / equipment requires independent verification by whom? I 2. Who determines whether a system i s critical or not when a jumper is .] to be installed? 1 i DUESTION 8.09 (1.00) l According to AD 1805.00, Procedure Preparation and Maintenance, what is the major difference between the uses of a procedure change and a procedure revision? OUESTION 8.10 (1.00) According to AD 1805.00 Procedure Preparation and Maintenance, what is the maximum period of time that temporary approval may be granted for a procedure? l (**+** CATEGORY 08 CONTINUED ON NEXT PAGE

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j 1 1 ? 1 ) e,__6DMINISIB6I12g_EBgCgpuggg,_QQNQ1IlgNg2_@NQ_LidlI@IlgNS PAGE if j s \\ i I QUESTION 8.11 (1.00) i During the performance of a Surveillance Test, a deficiency is noted. 1 According to AD 1838.02 Performance of Surveillance and Periodic Tests, under what 2 conditions is the Shift Supervisor permitted to l ~ allow the test to continue to completion? 1 QUESTION B.12 (1.00) l FILL IN THE BLANK j Tec. Spec. 4.0.2 States: { Each Surveillance Requirement shall be performed within the specified I time interval with:

a. A maximim allowable extension not to exceed ______ of the surveillance interval, and b.

A total maximum combined interval time for any 3 consecutive tests not to exceed ____ times the specified surveillance interval. QUESTION G.13 (1.00) l Pr ocedural l y, what is the LOWEST Emergency Classification Level requiring activation of the On-Site Emergency Organization? CilESTION B.14 (2.25) Per the D-D Emergency Plan 1. WHAT is the Primary f unction of the On-site Emergency Organization 2. LIST the five (5) groups that make up the On-site Emergency Organization. l (**+++ CATEGORY 08 CONTINUED ON NEXT PAGE ***++)

i 1 J eg__8DUIU1EIB8IlyE_EBQCEDUBgg3_CQNpillQUS _QUp_ Lid 11@IlgNg. PAGE 15 2 4 i g,, ,. :..q g QUESTION 8.15 (1.50) l {..;~t j. ;* s' es + n.. $ ; Per D-B Emergency Plan Sec 6.5.1, Emergency Personnel Exposure: 1. What are the exposure dose guidelines for: 1. Lifesaving action 2. Corrective action 2. The Shift Supervisor will initially have the authority to permit the Emergency Exposures. Once the Technical Support Center is activated, the responsibility is shifted to WHOM7 QUESTION 8.16 (1.50) Per D-B Emergency Plan LIST the three (3) designated areas where the emergency teams can be assembled and equipped. OUESTION 8.17 (.50) As a Shift Supervisor to what AD Procedure would you go to find H information on " Notification / Reporting Guidelines" to the NRC and l other agencies? QUESTION 8.18 (2.00) T.S 3.4.4 STATES: The pressurizer shall be OPERABLE with: 1. A steam bubble 2. A water level between 45 and 305 inches WHAT are the bases for each these i.S.s? i i 1 l l l (++*** END OF CATEGORY 08 **+**) 1 (********++*** END OF EXAMINATION ***************) l l l l

o PAGE 2C! Di__INEQBY_QE_U'JGLE88_E9EEB_ELONI_QEEBBIl0Ni_ELVIDEi_ BUD IHEBUQDYUBU1G5 ANSWERS -- DAVIS-BESSE -87/OS/17-SALYERS, G. M.01 ASTER CO.PJ ^ (2.00) ANSWER s

1. Actual critical position will be higher (0.25) due to xenon peak after the trip. (0.25)

2. Actual critical position will be lower (0.25) with less boron in the RCS the Control Rods will need to be inserted to compensate for the lower reactivity of the boron. (0.25)

3. Actual critical position will be higher (0.25) due to less excess reactivity remaining in the core at 200 EFPD. (0.25)

4. Actual critical position will be the same (0.25) since initial i

count rate has no effect on ECP (only power at which ECP occurs). (0.25) REFERENCE TECO PP 1103.08 and RTH-SRO series. 192OOOK103 192OOOK107 ...(KA*S) ANSWER 5.02 (1.00) 1. Axial peaking factor or Fz (0.5) i (Fz = maximum linear power density in rod / average linear power density in rod.)

2. Radial power factor or F delta h (0.5)

{ (Product of local rod power factor and fuel assembly radial peaking f actor) (local rod peaking factor = power in one rod / assembly average rod power) (assembly radial peaking factor = power in one assembly / average power in all assemblies.; I NOTE: any answer acceptable as long as the f actors Fr & F, delta h are i d en t ', f l ed. g gg /q g, -sqq 4,a., 9 '7 REFERENCE TECD HTT-SRO-010 192OO5K112 ...(KA'S)

PAGE 21 Ez__ISEQSY_QE_NWGLEBB_EQWEB_ELBUI_QBEBBI19Nt_ELVIQS _8NQ

+ IUEBdQQXNed1GE ANSWERS -- DAVIS-BESSE-

-97/OB/17-SALYERS, G. ANSWER 5.03' (1.50): 1

1. Control rods in a single group move together (with no individual

~ rod >6.5% f rom the group average position) (of 25 + Controlling groups are sequenced with appropriate overlap 2. or - 5%)

3. Controlling group insertion limits are not violated.

4. Axial power imbalance limits are met.

(any 3 8 0.5 ea.) REFERENCE TECOHhT-SRd-010pg12 192OO5K112 ...(KA'S) ANSWER 5.04 (1.25) 1. From decay 52 Te 135 B-(2 mi) 53 I 135 B-(6.6 hr) 54 Xe 135 B-(9.2 hr) 55 Cs 135 (0.5) Some f rom direct fission. Removal is from decay and burnup. (0,5) i Od *( 6.3x +.e* (,e 7 $ 0.25> 2. REFERENCE TECO RTH-SRO-07 192OO6K103 192OO6K104 ...(KA'S) i J _____-__________a

PAGE 22: 'Et__IUEQBY_DE_NWGLEBB.EQWEB_ELONI_DEEB8I1904_ELWIDE,_0ND .'INEBdQDYN9dIGE ANSWERS --' DAVIS-BESSE -87/OB/17-SALYERS, G. l ANSWER 5.05 (1.25) i

1. Combined effects of Mod. temp. coefficient, Fuel Temp. coefficient, and void-coefficient on reactivity.

(3 8. 25 ea. total O.75) ) Total. power def ect becomes more negative.f rom BOL to EOL since T

2. mod becomes more negative due to deboration.

(0.5) REFERENCE TECO RTH-SRC-OO4 pg 5&6*0NUM 192OO4K108 192OO4K113 ...(KA'S) l-ANSWER 5.b6' (1.50) Ensure that acceptable power distribution limits are maintained. l-1. (0.5) L

2. Ensure that the minimum, SHUTDOWN MARGIN is maintained.

(0.5) l I

3. Limit the potential effect of a rod ejection accident.

(0.5) 1 REFERENCE TECO T.S. pg B 3/4 1-3 192OO5K115 ...(KA'S) j i 1

' \\f a i .i

~ D-l r

Q.

THEORY _QE.,NUQLEBBJQWEB_fjL ANT _,,,QPERGTigh_,ELUlp@ i _6ND PAGE 2;1 IBE600DXN8dlGE. l e " ANSWERS -- DAVIS-CS E -87/OB/17-SALYERS,'G. i l ' ANSWER S. 07 ~. (2.00) '3. The, amount of reactivity-that.the reactor is or, instantaneously-could be made subcritical from-its present condition assuming.no'.' change in'.APSR position, and all control rods are fully inserted except for-the single rod of highest worth which.is assumed to be fu1]y withdrawn. ' 2. rod worth, reactivity of temp,' boron, xenon, fuel (0.25 es. total of 1'. 0) c tt 51vc W t e O 79 + Ig M-3/ 5- ..t REFERENCE .i TECD-RTR-OLC-023 192OO2K113 ...(KA'S) j ANSWER 5.08 (2.25) '1. Deta Core is the sum of all the Bi values for fission of a 1 particular fuel or the total fraction of delayed neutron's produced l by ission of p particular fuel C;- l /hr55*ME35'9 7-fr* 2I i 23f' 9 0 j ce n t - (Bi i= the fraction of total number of neutrons present in e reactor.which are produced by a given delayed neutron precursor.) I b ' ff ' "7I', f 2. B-U230 0.0156 i B-U235. O.0064- / B-Pu239 0.0021 3. Beta core decreases over core lif e due to; j U235 U239 Pu239 Power BOL 93% +- 3% 7% +- 3% 0% 4rection EOL 55% +- 5% 7% +- 3% 38% +- 5% (1.0) REFERENCE TECO RTR-OLC-010 TP 10.6 b 10.7, pg4 192OO3r107 ...(KA's) ( l.i. J l i- _ _ - _ - _ = _ -

n 'l., 1 , /- < ' lic THEORY _QE_NWGLE68_EQWEB_ELONT_QEE66TigNz_ELW1Q@i_609 1PAGE, 4:< l " *1 ~ cldEGUQEXd6dlGE h ANSWERS -- DAVIS-BESSE' -87/08/17-SALYERS,-:G. i ANSWER ~5.09. .(1.00) Given p = -6%' CR(1)~=~50 cps 285 cps CR(2) = 1 1 =. O. 943 OK i f estimated p = k-1 K = = 1 - p' '1 1.06 K 'CR 1-K 1 1 2 '50 2

, 4 CR

-1 .K 285-1 - 0.943 2 1 1 50(0.057) 0.99 .-____-____,. 1

K Therefore K

= ~1 235-2 2 p. ______.. _'O.99 - 'M-1 1 -J' ________. _o,og K O.99-or 1% shutdown (4 parts G O.25 ea.) i ( 1 REFERENCE TECD RTR-OLC-024 192OO3K102 ...(KA'S) g I ANSWER 5.10 '(2.00) 1 1

1. - 1. seven (0.25) rhodium emitters. (0.25)

'2. one (0.26) background' wire (detector) (0.25) j 3. one (0.25) thermocouple. (0.25)

2. When a rhodium atom in the emitter absorbs e neutr on, (0.1) it is transmitted.to rhodium-104,(0.1) which B-decays producing an electr on which produces e current flow,(0,1) thi s current + 1ow.i e carried by the lead wire to the Control Room.(0.1) The background l

( wire.is similar to a leadware and is used to account for,. gamma-induced' currents. (0.1) i 1 I l. (103Rh + N ---104Rh----194Fd 4 B-) E J .L__

g-y-- he ,:O p-

• ni.

1_ IdE98Y_9E_NUGLE68_EQWE6_ELONI_QEES9I19d2_ELLJ1DE2_9ND PAGE ;O! h X ISEBdQQyNSd1GS ANS'WERS:-- DAVIS-BECSE- -87/08/17-SALYERS,;G. r, .. g .c. REFERENCE

(

J. .1ECO PWR-DLC-038 .-015000K501 ...(KA'S) i ~ ANSWER 5.11 '41.00) I -l At2240 psia, hg.= 1115 BTU /lb.and at 20 psia, at saturation . conditions,.hg'= 1156 BTU /lb and hf.= 196 BTU /lb' 0.957 = 95.7 (+/-'2%)- - cal cul ates -(1115-196)' / (1156-196) = l If use.Mollier: 95% quality.(+/- 5%) REFERENCEf 4 TECO HTT-SRO-OO3-193OO3K125 ...' ( KA ' 5 ) NNSWER 5.12 '(1.50) 1 '1. RCS delta T has stabilized, does not exceed 50 deg. F (0.5) y 2. Verify heat removal from the OTSG's by observing turbine bypess l

val ve-position or atmospher ic vent val ve posi tion, (O.25) and auniliary.feedwater flow. (0.25) i 1

3. Incore thermocouple e temperatures stabili:e. (0.5) 4. The RCS is et least 50 deg. F subcooled per Tsat meters. (0.5) (cc pr rJrdai,s., e7 N,u, g f,; #) (any 3 C4 0.5 ea.) f REFERENCE 193OOOK122 ...(KA'6) r

1 Y s, [o ') ;_ (- . 9.;[hf__IhgG3YlQE_UgQ(E86_EQWE6_E(8NI_QEgB611QL_ELylQg,_@NQ. PAGE.2? Ji-: ' ' o THERdQDyN6MlQg ANSWERS -- : DAOI5-E4 ESSE ' -87/08/17-SALYERS,,G.. i '[ /.4 _ ANSWER. 5.13 .'( 1 ' so) -- /> b f if fl* . 1,. Increases (0.5) 2. As the speed of the pump increases the, pressure energy.is- ' tr ansf ere d to ' vel oci ty energy' at =. the eye' of the - pump. (O. 4) ThiC N ' causes ' a. decrease i n ' pressure at ~ the eye' of the pump, (0.4) 'thus getting closer to saturation.(0.2)- (Similar wording accepteble)'. ' ( 1. 0) 4 REFERENCE- ~TECO HTR-DLC-019 191004K106 ....(KA*S) 7 ANSWER: 5.14 (1.00) 3-P1-(N2/N1) = P2 3 60 (1800/1200) = P2 3 60-(1.5) = P2 l I 60 (3.375) = 202.5 kw . p. REFERENCE TECO HTR-OLC-019 191004K105 ...(KA*S) ANGWER 5.'15 (.7G) I i e. Reector Cool ant Hot Leg T emperaturre. (0.25) .j b.: Reactor Coolant Pressure (0.25) c. Reactor Coolent Flow (0.25) REFERENT TECO T.S. 3.2.5 19300EL105 ...(KA'S) a a 1______._ .._..m.._

PAGE E TEis.__ISEQSY_QE_NWGLE88_EQWEB_ELONI_QEE68119N4_ELW1DS _8N9 2 4, f,., . T d g B t]Q D Y N A M I C S - 0.- ' ANSWERS -- DAVIS-BESSE. -87/09/17-SALYERS, G.- l ) ' \\, _ i ' ' ANSWER. 5.16 (1.50) Dpri 1. .Mpri = --------- (O.75 f or equation) ~ . (hh - hc ) 2700 hW (3.41 btu /w) (0.5 for correct "h" determination)- .Mpri = -------------------- 622.8 - 552.3 btu /lbm- .I[.' 9207 E6 btu (0.25 f or calcul ation) Mpri'= -------------------- 70.5 btu /lbm c. + 10E6 ( <")# ), pj,.. [f / Mpri - =.130. 6 E6 ' 1 bm/hr. f -(1.5) p

1. I

// "/ REFERENCE c-1ECO HTR-OLC-023 d /C 195007K108 ....,(KA'S) b ~ bl

q; 1 g. j s h__IdE0BX_9E NWGLEeB_EQWEB_E6091_QCEBBIlgh_ELylph_09Q: PAGE 2

c. -

IMEBd99YN6digg. . ANSWERS - > DAVIS-BESSE- -87/OS/17-SALYERS,.-G. I LANSWER- '5.17-(2.25) r '1. 1. Formation of bubbles at imperfections on the surface of the j tube.(0.2) Heat.is concentrated at the. imperfection.-Bubbles are swept. away f rom the tube into cool ant. (0.1) The bubbles egitate .i the coolant, then collapse. (0.2) (As boiling rate. increases you reach bulk boiling. Here the steam bubbles do not collapse.) 2. The entire surface is covered by a film of? vapor,(0.3) which is a poort conductor of heat-and acts as more of an insultor than- ~ the liquid i t. repl aced. (0.2) (Energy in.this region is by rediation and convection). 3. Enough heat in added to completely vaporize all _ of the-liquid present.(0.3) The further addition of. heat causes the vapor to become superheated.(0.7) " p. / '/ ~~) / 2. 1. increases. 2. stays tho'same NI -< s t < t'a t

• 1 3.

decreases i 4 ' REFERENCE d l l

TECO PWR-OLC-026 L HTR-OLC-OOL 1

193OOSK103 ...(K4'S) ) '1 ANSWER 5.18 (1.00) 1.) Must have a heat sink (0.25) 2.) Must he.'e a heat source (0.25) 0.) An elevation difference between the heat source and the heat si ni:. 1 (0.25) l 4.) Coupling of a = b; cal " liquid between the heat sink and the heet l source. (0.25) on p ./ c s,. . / (d j b // /[, # r

• 4

!* b ' REf*ERENCC t 1ECU HTR-OLC-11 , l /, 193OOGK121 ...(KA'S) ' ' ^ }' _[ J/ 1 I. m___.____-_*_-_a

w k :.. ;:p s. c .. i,, e PAGE ?- .',}'- m.l

n __CLGNI_EYSIEdE_QEE1 Qui _QQNISQ63_6NQ_10@I6QUENIQIlgU

[ ANSWERS - IDAVIS-BESS'E' -87/OS/17-SALYERS,-G. ANSWER.

6. Oi!

(1.00) N

• la.'SFRCS CHANNEL 1------b:

( 1. O ) -6 M.SFRCS CHANNEL 4------i \\ g 4 '5GL A C 2--------------qi ( jjf,ggtIt3-_____________o_ L y, i REFERENCE -1 .TECO PWR-OLC.-030 DAVIS BESSE: TRAINING:INFORMATION. MANUAL VOLUME 5, SFRCS* iO61000k201 '...(KA'S) 'Ai45WEFi - 6.02 (1.25) ~ 1..' Low Main. Steam Pressure -612 psi g c),;,g,/#.

2. H2gh'SG/Foedwater D/P -177 psid

.,,h.

3. High SG. Level -280"

?O4 g,* g_y, 26.5" 4.1-low SG Leve) 5. Loss.of Four (4) RCP's l l '(0.15 'f or each item; O.10 for each setpoint) REFERENCE TECO PWR-OLC-030 1 DAVID BESSE TRAINING INFORMATION MANUAL VOLUME 5, SFRCS 061000U402 ...(KA'S) i i l I i-t 02-. L._ - -

_,_c_7___ t e_ _E66WI_EXEIEd@_pg@l@U3_QOylBQLt_889_IN@lBUDEUIeI1QN PAGE 3i i ANSWERS -- DAVIS-BESSE -07/OG/17-SALYERS, G. ANSWER 6.03 (2.00) { i 1. In the AUTO ESSENTIAL position (0.5) OSTG 1evel is controlled at 49 j inches on the startup range when a AFW pump is aligned to its own generator. (0.25) l If there is an SFAS Level 2, present, level in the OTSG is controlled at 124" (0.25) When an AFW pump is aligned to the opposite OTSG, level is controlled to 55" or 130" if an SFAS Level 2 is present. (0.25)

2. In the MANUAL posi tion, (0.5) the operator controls the speed of the AFW pump to control OSTG 1evel. (0.25) i REFERENCE TECO PWR-OLC-030 061000K411

...(KA'S) ~ p' 1 '/ i ANSWER 6.04 (1.00) m lYli ( * *- ) I 1. (656' El.) - Top of 41 Steam Generator secondary shield' wall (Normal for 1-1) 2. (603' Ei.) - Personnel Air Lock area (Normal f or 1-2) l .r 3. (017' El.) - Top of the CTMT Dame 4 4. (656' El.) - Top of H2 Steam Generator secondary shield h,e -, ) i well (4 6 0.25 ea) i l ) REFERENCE l DAVIS DESSE TRAINING INFORMATION MANUAL VOLUME 7, HYDROGEN REMOVAL SYSTEM CTf1T. 028000A403 ...(KA'S) l l l l l l .---____--Q

( 6 __ELONI_MSIGMg_pggigh_QgUI6QLx_6NQ_lNSIB(!dENI6110N PADE 3 9 ANSWERS -- DAVIS-BESSE -87/08/17-SALYERS, G. ANSWER 6.05 -(1.00) p Gased from CTMT are directed to the Hydrogen Recombiner heaterc via i the Hydrogen Purge line. (0,5)The gases are returned to CTMT from the Hydrogen Recombiner through the 1-2 Hydrogen Dilution Blower discharge line. (0.5) REFERENCE DAVIG BESSE TRAINING INFORMAT]ON MANUAL VOLUME 7, HYDROGEN REMOVAL SYSTEM, CTMT 02OOOOK101 ...(LA'S) I ANSWER 6.06 (2.00) 1.

1. -Turbi ne t ri p signal -Emergency trip system oil pressure. (on the main stop valves less than 275 psig in the turbine EHC).

(0.5) 2. Main feedwater pump trip signal - control oil pressure (l ess then 75 psig on both pumps) (0.5) 3. Out of core flux signal - from RPS (0.5) ///['/'/C //0* (0.5) 2. Loss of all RCP 's evt REFERENCE '/ f i TECO PWR-OLC-040 SP 1105.21 DAVIS BESSE TRAINING INFORMATIO MANUAL VOLLIME 6, ARTS 012OOOK402 ...(KA'S) ANSWER 6.07 (1.00) i ' C 'l/ !~ / g a. Fal se /', (0.25 ~ '/ ^ s b. (One is located on each side of the under-voltage coil) opening the contacts deenergines the UV coil (0.5) f or t h e B R:: 1 rip breM er. (0.25)

g.... .o [. PAGE 5-b_E(ANT ' SYSTEMg_DggigN _QQNTBQL _ANQ_INSTBl_.lMENIATiQN 3 3 y ANSWERS - ' DAVIS-BESSE -87/,08/17-SALYERS, G. 1 I T REFERENCE TELO PWR-OLC-040 012OOOK105- ....(KA'S) ( , ANSWER-6'.00. .t (1.50) ' 1. Prevents backing. water into the HP turbine, which.could cause turbine' blade damage. (0.5) oc : ecitycver g4e & c p we 0,44 .(Moisture f rchm the steam enters the MSR from the exhaust of the 2. I HP turbine.) As.the steam enters it first. passes over chevron

plates that remove moisture (0. 25) (by causing ' abrupt changes i n.

direction.) Thi s moi sture drains down f r om -the plates-in the MSR .into the-MSR drsin tank.(0.25)

3. Feedwater Heater 5-1 or 2 4.' Condenser REFERENCE f

.FtilD. MOOS 1 TECCi PWR-OLC-02G 039000K105 ...(KA'S) I ANSWER 6.09 (1.25) el. Cores, Support Shield I '2. The Internal Vent Val ves prevent a pressure imbalance which could interf er with emergency core cooling f ollowing a LOCA. (under LOCA conditions'due to a cold leg rupture, the pressure within the CORE SUPPORT ASSEMBLY will be higher than the pressure within the annulus. The Int ernal Vent Val ves would relieve the inter nal pressure out the cold 3eg break.)

3. start to open at 0.154 (4/

• I i'f

4. f ully open at 0.3H g,g g j

REFERCNCE TECO PWR-DLC-031 1)(;VIS IICESE TRAINING INFORMATION MANilAL VOL.1 REACTPR VESLE; .b

.1 4 ' ij e, q l], b.,__E!=6BI SXSIEti3 DE@jGhi_CgBIBL,8ND_1NSIBWEUISIlOB . PAGE'3.h ' ANSWERS - ' DAVIS-NESSE ! 87/08/17-SALYERS,-G.. E l L. ' AND INTERNALS. H CO2OOOK613-(KA'S). l l l S f lf l. l/ ANSWER 6.10' (1.50) 10.

y 1.

SFAS. Level'1. Actuation

2. Station Vent High Radiation Level 3.

High. Chlorine (5 ppm G...ventil ati on inlet or blockhouse)' 4. Lor s of, control.: ai r.(<.75 psi g ) L

5...L~oss of; control power j

.6. Free:e Stats (35.deg.F'O outlet of heating coil) ,7..Stnoke Stats (3% smoke density e outlet of Air. Handling' l Unit'or inlet to the Return Air Fan)- '8. Fire Stats..(135 deg. F @ outlet'of Air Handling. Unit or 1651deg.F G' inlet to the Return' Air Fan) 1%.hou ,t- @ b c /A{ (5 @ O.3 ea)- ' REFERENCE .TECD PWR-OLC-OO9

I)AVIH BESSE 1 RAINING INFORMAT10N MANUAL VOL. 8 CONTROL ROOM

~ VENTILATION .000060EL30 (KA'S) o t f 4 a 2___--..

o 1; Les__ELONI_@ygIgd@_Qg@l@U3_QQN16Q63_6NQ,1NS16gdgNI@llQd. ~PAGE 3- ' ' ~ ANSWERS - JDAVIS-BCSSEJ .-87/08/17-SALYERS, G.' l ' ANSWER. 6.11 . (3. 00) ' 1.$High'RC Bldg [Precs. a..Setpoint'<:or-= 4psig (0.25) b. Reason: Provides positive assurance that a reactor trip will occur in the unlikely event of a steamline failure'in the containment or a LOCA even in the absence of c RC Low Press Trip. (0.75) 2. RC:High Temp a. Setpoi nt <.or = 618.deg F. (0.25) 6.L' Reason: _(1) Prevents the reactor outlet temperature from exceeding design limits (650 dag F.) or (2) Backup trip for.all power excursion transients. (0.75) 3. RC High Press 'i i .1. Setpoint < 2300 poig (0.25) O.-Reasons. j 1. Backup to the high flux trip or 2. To maintain.the system prer.sure below the safety limit of 2750 psig for any design transient. l or 3. Provide. start-up accident protection from low power or ,4. Slow reactivity addition from high power to prevent' exceedi ng 2750 puig sa4ety' limit (0.75) i REFERENCE T ECD ' PWR -OLC-039 012OOOV402 ...(KA'9) u_-~.____..__-._

hu._E66U1_EYEIEdS_DE@l@ h _GQNIBQ62_6ND_lM@IBLJdENIGIl99 PAGE '.'d ANSUERS -- DAVIS-BESSE -87/08/17-SALYERS, G. ANSWER 6.12 (1.00) I 1 f vifDb.d y&nl bc u i 2. P.ower / Imbalance / Flow Trip /" * EO l ^'/' 2. Power / Pump Trip / i' / 3. Press.ure / temperature Trip / $/ 4. Low Pressure Trip i REFERENCE TECO PWR-OLC-039 SP 1105.02 0120001:604 ... O.: A ' S ) i l 1 l {'.

b' r ia 3,_ b ] 6f__CL6Sl_EYEIEdE_DECl@y,_QQUISQL,_69D_1ggI6ggggI611gg PAgt 3, 1 ? ANSWERS'-- DAVI5-EESSE -87/OB/17-SALYERS, G. -l 1 q ANSWER-l 6'.13 ' (2. 00 ) -i e i ] i g. 1.'The power range detectors 5-8 are in 6,eries with the; sour ce and intermediate range--dutectors 1-4 (0.5) l 2. For Power range 5-8 it must be 5 Pr 6 or 7't< 8 arrangement. (seri es ' 8< parall el ) (0.75) J J 3. NIs 1-4'must' be in parallel (0. 5 ) - 4. >10% flun power range (0.25) CONTACT POSITION NOT IMPORTANT -1 1 ZDDDDDDDDDDDDDDDDDDDDDDDDDDDDDDDD7 3 3 i 3 ZDDDDDDDDDDDDDDADDDDDDDDDDDDD'? 3-DDADD DDAD-3 NI 5 DDBDD 'NI 7 DDDD 3 3 FLUX > 10% 3 ZDDDDDDDDDDADDDDDDDD7 DDADD 'DDAD' I

3. Inhibit Control 3 NI 6.DDDDD-NI B DDBD I

3 Rod Withdrawal 3 .3 3 3 3 GDDDDDDDDDDDDDDDBDDDDDDDDDDDDY GDDDDDDDDDBDDDDDDDDDY 3 3 3 3 ZDDDDDDDBDDDDDDADDBDDDDDDDDD7 3 3 3 3 3 3 DDAD DAD DDADD DDAD 3 NI 1 DDBD NI 2DBD NI 3DDBDD N1 4DDBD i 3 3 3 3 3 l 3 3 3 3 3 3 GDDDDDDDADDDDDDDDDADDDDDDDDDY 1 3 3 GDDDDDDDDDDDDDDDDDDDDDDDDDDDDDDDDDDY REFERENC'E TECO SP 110b.V2 OO1000K403 ...(RA'S) 1 )

7 fa__t L OUI _ E X E1E U E _ D E E1E h _ G 991bO Li_6N O _lN EI6 W Ugyl@llg N FAGE 0; AN5WERS -- DAVIS-BESSE -87/OB/17-SALYERS, G. 1 ANDWEh c.14 (1.00) 1 1. Cl een Weste Sysstem 2. closed REFERENCE TECO PWF(-OLC-OO4 SP 1104.27 071000A429 ...(KA'S) ANSWER c>.15 (1.00) A RCS F' low error of 10% between loops will cause the auto transfer of Tave t the loop with the highest f l ow. O /L 'c 5 5 el tj i' 1 ()Mco. f f y. 37 (1*O) REFERENCE TECD PWR-DLC-O!D DAVIS BESSE TRAINING INFORMATION MANUAL VOL.5 NNI 016000K101 ... (I.: A ' S ) ANSWER 6.16 (1.00) Cabinet X REFERENCE ST'1105. 06 lEt/O PWR-OLC-075 0160006'.201 ... (6' A ' S)

m.. \\ '1: e..., L y%'* PAGE .)_ ( s dr._%6BlEX@lEUE_EEEI_EUA_COUIbQha AND ' 1NST RUMENT AT it']U. '- 3! L. ti s ANSWERS -- DAVIS-BESSE - , L -87/08/17-SALYERS, G.. t e ANSWER 6.17 (1.50) 4,fM \\a t 1 j ],(' 1 ITECO.PWR-OLC-075 , 045000K410 ...(KA'S) . ANSWER 6.20-(.50) . Stator Coolant : Runback ' REFERENCE DAVID BESSE TRAINING INFORMATION MANUAL VOL.3 EHC L PWR-OLC O75 ~ 045000kA12: ...U:A'S) 'ANSWCR -. 6. 21 (2. 00). g- -1. Coritrel Rod Drive Mech. l 2. Reactor Coolant Pumps 1 i 3. Reector Cool ant Motors

1. d4's 'A g', g,g.

4.;Make Up Pumps D. E me-r g. Instr.1 Air Compressor g) 4, _13 A ) j - 6. - Le down Coopers if, )q, my y f e,' fy I4-G/ l . REFERENCE i W -GOP-OLC-OO1 Attathment 31 [ l l"- AB1203.31 IQ4 .J' l-/a RO 3.6 SRD 3.5 l GOOO26G010 OOeOOOc.01 ...n#S, e H C-i

l LZz__ PROCEDURES - NQRd@Lt_AENQBdALx_EMERGENQX_AND PAGE 44 Se91969GICeL_gDUI696 ANSWERS -- DAVIS-BESSE -87/08/17-SALYERS, G. / C) ANSWER 7.01 RCP's must be tripped within two minutes after losing SCM to prevent the RCS.from reaching a significant void fraction (0.5) such that the core would be uncovered if the RCPs were tripped at a later time. )on the s all br k anafysisy't (1.0) Th two finutes fs based 2 earliest that 4 signifdcent (70 %) void f action ould 4ccur/ O..S) % ota tmg '/ REFERENCE GDP-OLC-OO3 TECHNICAL BASES DOCUMENT pg IV.A-2 OOOO74EK30 ...(KA'S) ANSWER 7.02 (1.50) a. Fails open. b. Fei l s as, i s. c. Fails hal f open. REFERENCE EP 1202.01 GDP-OLC-OO3 Attachment O2 pg 6 (Supplemental Actionc) OOOO28EU20 ...(KA'S) { l i i

• e 1

l l w-_ -_ --

r PAGE 4.t 2t__Eb9GGUU6Ek_:_UQBd662_6pNQSd662_EdEE@GUGY_@UD BOD 1969G1G66_G9BI696 ANSWERS -- DAVIS-BESSE -87/OS/17-SALYERS, G. l l ANSWER 7.03 (2.00) 1. 1. 100 deg. F /AV 2. 50 deg. F o& [ l 2. 2. Average SG Shell temperature - 1 cold (ten si l e) 2. T hot - Average SG Shell Temperature (compressive) A /

1. 1. " 7 REFERENCE AB 1203.40 step 3.21 TECHNICAL BASES DOCUMENT SEC. III.G-19 i

GOP-OLC-OO3 ~ 0.5010 GOOD ...G A'S) ANSWER 7.04 (.50) i True (0.5) REFERENCE-TECHNICAL BASES DOCUMENT SEC IV.D-1 017020A101 ...(KA'S) i

Zt__CB99EDUSEE_:_UQSd861_ GEN 9BdeL2_EDEE@ENGY_6NQ: PAGE: C E09196991G86_G9NISQL ~~ ' ANSWERS -- DAVIS-BESSE '-87/08/17-SALYERS, G. g l 4 l l . ANSWER 7.05 (2.00) l. l 1..The'20 minute delay provides reasonable assurance that the primary system will:notirepressurize and' result in.a loss of LPI, flow.: (The decay heat l evel i s low enough so as not to_ boil the coolant causing a repressur2:ation.)

2. N.9 minimum; required LPI flow rate is used to ensure that the injection flow can remove decay heat afterothe HPI is stopped.

5. The LPI - flow rate is required in each line to assure that at least-the minimum required LPI. flow is reaching the RV (in the event that a break exists in one of the LPI/CF lines which could prevent LP1 water.from reaching.the RV through one of the lines.) REFERENCE TECHNICAL BASEE DOCUMEMT pg IV.B-12 GOF-OLC-DO3 000031 El:.31 ...(KA'S) ANSWER-7.06 (1.00) The average suction temperature of the. running CTMT Air Coolers. REFERENCE 1 GDP-OLC-OO3 ATTACHMENT 1 pg 9 OOOO11EK2O O22OOOA405' ...(KA'S) 1 -) 1 l ftNSWER 7.07 (1.00) 75 psig REFERENCE ) 1 AB 1203.36 LDSS C'F INSTRUMElJT AIR I:/a RO 3.6 SRO 4.2 OOOO65EATO ...O'A'S) ] l

hl r i Zr w;.EB9G699 BEE _~_d9606!=t_61!d9Bd662_E_DEEQENGy,,99p - PAGE 'C ,4 e' E E6D1969@lG66_G9dIB06. ANSWERS -- DAVIS-BESSE -07/OB/17-SALYERS, G. ANSWER: 7.08 (i.00) f(With one pump' running in a'. loop,'and. operating.on cooling tower bypass,)'if.the circulating pump'should trip, the condenser woul d parti a11 y '. (0. 5) drainL(due.to the height difference.) If the. complimentary.. pump were started it would cause a water hammer (due to ~ the water fl ow accelerating to the' condenser,) and. damage the tube sheet. (0.5) '(simi 1 ar.' wording acceptable) REFERENCE AB 1203.24 CIRCULATING. WATER PUMP TRIP / CIRCULATING WATER SYSTEM' RUPTURE 'GOP-OLC-OO1 Attachment 24 ~ SP-1104.09 CIRCULATING WATER SYSTEM. lu a. RD 2.5 SRO 2.7 075000A202 .. 0;A'S) ANSWLR 7.09 (1.50) For leaks greater than 50 gpni you close the Turbine Bypass Valve and the Atmospheric Vent Valve on the affected OTSG when the RCS temperature is <410 deg.F and pressure <1000 psig. If the l eal.: is less than 50 gpm then.you steam both OTSGs all the way down. REFERENCE AD 1203.40 GOP-OLC-OO1 ATTACHt1ENT 40 l./a RO 4.2 SRO 4.4 OO'.iu37 CKP O .. 0:A '5) E __J

30 m .O( LZh__CB9GEDWBEE_:_NQBd@62_6ENQBU662_EdES@EUQX_6SD PAGE 4 690196001G66_GQNI696-ANSWERS.- ' DAVIS-BESSE -87/08/17-SALYERS, G. j i 1 . ANSWER 7.-10 (1.50)' I In analy:e,-the monitors are set to monitor 7-N-16 gammas (a very high ~ energy l gamma). (0.5) When the reactor ~is S/D, 7-N-16 production is . almost nonexistent and the 7-N-16' inventory decays rapidly due to short' half life.J(0.5) To restore'the monitors to service, select'the . gross mode at the monitor. (0. 5)

REFERENCE TECO Duestion Bank $153 (secti on 7)

EP 2202.01 Tube Rupture Step 8.3.2 SP 1105.07 see 1.5 pg 3 k/a RO 2.4 -SRO 2.4 ~ ~OOOO37EK2O ...(KA'S) ANSWER 7.11 (2.00) 1. The f uel pin iri compression limit is designed to prevent = ire hydriding of the fuel cladding; in the radial direction which can j weal:en the cl adding. (aligned in the radial direction allows the platlets to slide easy on adjacent moleucles.irt the crystalline lattice.) (1.0) 2. (Below 42D deg. F'rirc hydriding is not a concern.) Due to the mass flow rate di f f erence. between f orced ' flow and natural circulation, (0.5) the delta T between-the coolant and the clad will be higher 4 or the ' natur al circulation flow, therefore, the clad temperature will be higher f or the same indicated T' hot. (0. 5 ) (f orced / natural flow) (break point for forced flow 415 dog., for natural'cir. 380 deg) REFERENCE 4 'l TECilNICAL DAEEL DOCUMENT PAGE III-E-25 q k/a RD 0.8.SRO 3.9 OO2020 GOO 1 ... (1: A

• S i 1

I

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$4 4 1 y 42.t. E69CEEk!6EE_i_NQBd6LA_6kUQSd66t_EdgB@gugy_@ND ~PAGE e, ' f [%n - p B69196991G66_CQNIB96 ,: ANSWERS.-- DAVIS-BESSE- -87/08/17-SALYERS, G. L t o _ b / l l.r [,it: . ANSWER 7.12 (3.00). L Lj .1. Acoustic Monitors located on.the ~ PORV discharge pipe,.and provicW the followinglinformation, h Flow.

(Ranges; O.0-1.O) g

. % )- (50% open corresponds-to a 0.25" flow indication) L. " Position: ( (y . ind. (Open W/: Flow.>O.25) Open /Cl os ed (Closed W/ Flow <O.25)

2. HIS RC2 (located' on Panel' C5705)

The posi tion of the PORV Solenoid / Pilot Valve Lever-is indicated by the red and green lights. 3. ILJ.ftC2-6/ Blue Li ght '.(located on Panel C5705) This blue light i9 lit when 125 VDC power-(DBF-30) is available to the FORV pilot solenoid and. PORV cor.tr al ' ci rcui try. REFERENCE ~ ATi 1203.19 Pr essur i n er System Abnormal Opration TECO GOP-DLC-001 ATT.19 k/a RO 3.9 'SRO 4.1 OOOOOGEA20 ...(KA'S) .i 1 l ANSWLa 7.13 (1.50) 1 1. 1. When directed by another procedure 2. When in the judgment of the operator, plant conditions indicate j this procedure should be implemented. 1 2.. Level one on CTril RAD REFERENCE l I EP 1202.01 l k/a RO 4.1 SRO 4.3 000007G011 ...(LA'5) l I t. i

l, 2 }; }' ct; e l i, ' #.?i _PBQQgDURgg_ _NQRMAL-dA@NQBMAE_EMERQgNgy_AND-lFAGE > 4 c' : h H SBD196901G06_G9 NIB 96~ m ANSWERST-- DAV15-BESSE-- -87/08/17-SALYERS, G. 4 1 Io AfdSWER' 7.14' ' ( 2. 00 ) 1: 1'.'+ or 25 (0,5) 2.- 15Y

(O. 5) -

3. 15 ( S. 5) .4.i The promptireactivity-change associated with the movement of the ~ r ad s.. - (0. 5) REFEREi 'TECO.PP~ 1103.08, Approach To Criticality k/a' RO 3.3 / 3.4 SRO 3.5 / 3.9 L.OO1000G010 OO1000G011 ...(UA'S) ANSWER ' 7.15 ' (2.00) 1. Tr uc 2. True 3. True 4.:True REFERENCE AD 1203.32 STEAM GENERATOR FEEDWATER CHEMISTRY OUT OF SPEC TECO GOP-OLC-OO1 ATT. 32 L/a RO 2.7 / 3.4 SRO 2.9 / 3.6 056020G015 OS9000G015 ...(KA*S) I mau-----_----w-_L-.

b; Si__GRdlu1EISBIIVE _EBgcggggEg,_ggyn111ggg,_eNQ_(ldlI811gdg - PAGE. 4 ~"~ . NSWERS - DAVIS-BESSE- .-87/OB/17-SALYERS, G. -l ANSWER 8.01 (1.00) l'. hydrogen ( < 4 */. ), (0.5) oxygen. ( < 2% ) - (0. 5 ) o' 1-REFERENCC' TECO T.S. 3.11.2.5 k/a RO 2.4 SRO.3.1 071000G011 ...- (KA ' S ) { ANSWER 8.02 .(2.50) (any 5 of. tbe 6) \\ ~ y }(,,,_ ,,,,,,,,,, g fr f -r 7. *5 / . C : - ^. ;t b = ~m'=& s / .j 2..Cannot be located 3 3. Control rod is misaligned with its group. average by more than 9 i nche s :1. 4, 7 'd #N 7 7 -3 / 4. Control rod does'not meet the enercise requirements 5.' Control rod does not meet the red trip. insertion times 6. Control rod does not meet the rod program verification ~ (5 G O.5 ea) ' 7. [~ g/ .no4N..w , kNfPoy?W kd E.g' 7.g, -. f u -t o f REFERENCE y TECO T.S. 3.1.3 sec. 1 t I t/a-RO 3.4 SRO 3.9 Op1050G011 ... ( K A S ) ~ ..N / y('1.00) L f

6. '~ w n ' r, l

O ANbWER ,,g N [ / f ' I (,N g 7>' l ( /' l RE F EREICli TECD AD 1000.04 Post Accident Radiological Sampling and Anel ysi s " l OOOO76EK30 ...O'A'S) d__m_-_

a -e v a.w o .Qi__8dd101E186I1YE_EB9CEDUBES2_G9ND1119NSz_6ND_ Lid 116Il0NS PAGEa'4( ,...e

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G. l .' ANSWERS.-- DAVIS-DESSEE -87/ 08/1'7-SALYERS, i '.7!,.',_ g i l-l: ANSWER-8.04, -(1.50) p .1'. taking.'the reactor, critical. (0.5) .2. planned power change > 20 MWe/ min ~ (0.5) 3. pl anned unusual or abnormal.- events ( 0. 5 ) ' REFERENCE i ADM-E90-OO1 J.1 : .TECO AD 1839.04 k/a RO 2.5 SRO 3.9 194001A103 ...(MA'S) ' ANSWER D.05 .(1.50) b _ (tt..f 1.

1. - Notif y Assi stant' Plant Manager-Operations (0.5)

Notify NRC (within.one hour) (0. 5 ) 2.b"sureoperation is not resumed until approved y the NRC. (0,5) r% SM re-Vir (. /E!rinpG, fh r can ,i RE E . f,, [ o,,, ,,,, /g,, g (,,(, g,,. , gp3 gcf4, ADM-SRO-OO1.01 I.1 AD 1839 k /c RO 2.5' SRO 3.4 194001A103 ...(RA'S) ANbWER 8.06 (1.50) M f,</ o/T O ,d (0.5) 1. ST bchedule t/ (0.5) 2. El Al er t Report pf $. ~ y (- (0.5) 3. Critacal ST Report REFERENCE-ADM-SNO-OO4.01 B.1 AD 1 LC9, 05 1__

t-h,___6Dt!1NIEIBOIISE_EBQGEROBESt_G9UD1I1QNQx_6MQ_61hlI@llgNE PAGE ti ' u. J' s -ANSWERSL- : DAVIS-BESSE~ -87/OS/17-SALYERS, G. I-l k/a RO.2.5. SRO-3.4 194001A103 ...(KA'S) j, ANSWER 8.07 (1.50)' gphfitfb'h . i, i 1 b (0.5) fl 1, Ic'. hh/ f been completed. Id NN M (0.5) holding-clearancehavereleasedclearance.(o 1. Wor k tiat, d 2.,All personnel

3. The SS/ Assistant SS have determined-that the equipment' is ready for ggg o,, p(/;

(0.q) Qgy,,phe.q'ce,qWffj service. k,nuhnd) (;cp;9 REFERENCE ADM-SRO-005.01 TECO AD 1803.00 6/a RD 2.5.SRO:3.4 194003.A102 ...(KA'S) ANSWER 8.08 (1.50)

1. A qualified Journeymen or above.

(0.75) 2.-The DBTS (Davis-esse Tagging Supervisor) (0.75) C> ll. 5 A # T 5Y Ca.V!SctG f f M/ REFERENCE ADM-SRO-OO6 B.1 TECO AD 1823.00.17. 6 /a RO 2.5 SRO 3.4 194003 Allo- ...(KA*S)

L33__8Dd1GlEIEGIIME_EBQGEggBggz_GQUQlIlQUg2_QUQ_LidlI611gNE PAGE o +.!.* ' ANSWERS -- DAu15-BESSE -87/OS/17-SALYERS, G. 1 -' l ANSWER 8.09 (1.00) Procedure changes 5 are not allowed to alter the intent of the procedure while a revision is permitted to alter the intent. (1.0) REFERENCE ADM-SRO-012.01

AD1005.OO k/a RO 2.5 SRO 3.4 194001A103

...(KA'S) ANSWER O.10 (1.00) 14 days (1.0) REFERENCE ADM-SRO-012.01 I AD1805.00 k/a RO 2.5 SRO 3.4 194001A103 ...(KA*S) ANSWER 8.11 (1.00) 1. The deficiency i s minor (0.5) 2. The deficiency does not affect operability of the equipment (0.5) i l REFERENCE ADM--GR O-013. 02 i Ar.1638. 02 I 1 6/e RO 2.5 SRO ?. 4 194 001 A 10'.. ...(KA'G) I i

i., L.., Es _6Udidl@lS611ME_C69GEEWBEE2_G98911190EA_6MD_LidlTAllQN@ '.PAGE. 5: R " '

v.

~. - i ANSWERS --- DAVIS-DESSE .-87/08/17-SALYERS, G. A 1 I., i

ANElWER/

-8.12 (1.00)- L l; f:

c - l

a. 25/.- o 4

/'a 78 / (o.5) .:k < b. 3.25 (0.5) I REFERENCE '- TECO TECHNICAL', SPECIFICATION 3/4.O.2 .h/e RO 2.5 'SRO 3.4 194001A103 '...(KA'S)' + .' ANSWER G.13 (1.00) Alert _ - ( IL. 0 ) REFrIRENCE - D--D '"E mergenc y Pl an Sec 5.3.2 k/a RO 2.5 CRO 3.4 194001A116 ... (KA ' S): I ] ANSWER G.14-(2. 25) - f

1. To relieve' the Control Room Staff of those duties and j

responsibilities not directly'related to the operations of..the-primary and secondary p1 ant systems. .( 1. 0 )~ 2. 1.. Control Room' (0.25) 2. Technics 1 Support Center (0.25) 3. Operations Support Center (0.2S) 4.s Hea2 th Physics Monitoring Room (0.25) 5. Emergency-Control. Center (0.25) REFERENCE D-P Emergency Plan Sec. 5.3 6:/a'RO 2.5. SRO-3.4 194001A116 ...(KA'S) i a( --s_----,_n-_1.w,s--,---a.---D .a

r 9 ,. ~ e s JD2.deDd181EI6911YE_EB9GEQW6E h_G9091I19 E _00Q_ Lid 11SI! M a

0.

-87/08/17-SALYERS, G. ' ANSWERS --- DAV2'S-BESSE. i F- .AN5WER 8.15. ( 1. 50 ) ~ .(0.5)

3,.g,.'75 rem (0.5) l

' 2. 25 rem. (O. Emergency Piant Manager.,f,..g ? g 2. . REFERENCE D-D Emergency Plan See:6.5.1 =k/a RO 2.5' SRO 3.4-

194001A116' 194001K103

...(KA'5) "' O.16 (1.50) ANSWER -(E d (0.5) H ' i.. Turbine Dect: (O.5) 2, Health Physicu Monitor Room

3. Radi ol ogi c al Testing Leboratory' (0.5)

. o p r/ i 'ic e a. { S j r,oes T r & i. N / k 3 0 e M llc CE T REFERENCE D-ht. Emergency Plan Sec-6.3.2 -k/a RO 2.5 SRO 3.4: 194001 A116'. ...(KA'S) 'ANSWCR -8.17. (.50) ~" CONTROL OF CONDITIONS ADVERSE TO QUALITY" .AD'ibO7for REFERENCE AD-1007' k/a RO 2.5 SRO.3.4 194001 A10'?. ... WA'S) =

Dz__02d1NISIB611W ES9G i ' ANSWER 5L. -~ DAVIS-BEhSEL ,EQQBG@2_GQUQlllQU 4 $2_6NQ_(ldll611Q6@L e a- ,01/08/17

. ) -

-SALYERS, G., PAGE-e

ANSWER

'.a j. i= ,4 O.18 ,- l (, (2.00) , A steam bubble in the 1bydrauli presnure'cally solid system urizer press surge .reli eve RCS.pr s during oper (0 5)- ent.ures'that !pr. we ' l ow l evel essure during allation, and - i s the RCS is capab and the s t eam'l e notIa l bubble functionof accomm Syste r esctor coolantis based on providesign transient event a limit a ectuate-the Reactor s. (0 5) Protection Systsystem low press. enough water s-to ding m vol ume.. (0 5) The high level t-an si en t.to. prevent tr ure 1 volume to em or the Safety Featcondi tion that would (any 'such wo(0 5). a pressuriz l itni t is based. on provi rding) er high~ level ding' enough steure Ac REFERENCE as a result of any am TECO:T.S'3/4.4 4 L/a RO 2.6 -'011000G00 "SRO 3.7 6

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