ML103050185
| ML103050185 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 07/16/2010 |
| From: | Division of Nuclear Materials Safety II |
| To: | |
| References | |
| 50-338/10-301, 50-339/10-301 | |
| Download: ML103050185 (570) | |
Text
{{#Wiki_filter:North Anna 2010 Initial Exam Draft RO
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 I. OO1AK1.03 001/NEW//H/3/3.9/4.0/1/NO Given the following conditions:
- Unit 1 is at MOL, starting up following a forced outage.
- The crew has just completed taking critical data.
- The OATC pulls rods to establish a positive startup rate.
When the OATC releases the IN-HOLD-OUT switch, rods continue to step outward. Assuming NO operator actions are taken, which ONE of the following identifies the plant response to this event? Power will increase to The power increase is terminated by the A. 30%; Intermediate Range rod stop signal. B. 30%; Power Range rod stop signal. C. 25%; Intermediate Range reactor trip signal. D. 25%; Power Range reactor trip signal.
- a. Incorrect. Plausible since there is an IR rod stop at this power level, however the power range power neutron flux low setpoint which is blocked during a normal S/U and power ascension is still active based on these plant conditions so it will be encountered first.
- b. Incorrect. Plausible since there is a rod stop at this power level, however as noted above it is an IR rod stop (the PR rod stop is 103%) and also as noted above the power range power neutron flux low setpoint (25%) which is blocked during a normal S/U and power ascension is still active.
- c. Incorrect. Plausible since there is a trip at this power level but as noted above it is from the PR, not the IR; the IR trip is at 35% not 25%.
- d. Correct. As noted above this trip is blocked during a controlled power ascension, however this is a manual block unlike P-8 which occurs automatically. Thus based on these plant conditions the PR Flux Trip (Low Setpoint) will be encountered first, however if the candidate has the incorrect mental model they may eliminate this correct answer from the selection of choices.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Continuous Rod Withdrawal Knowledge of the operational implications of the following concepts as they apply to Continuous Rod Withdrawal: (CFR 41.8 /41.10/45.3) Relationship of reactivity and reactor power to rod movement Tier: 1 Group: 2 Technical
Reference:
dwg 5655D33 sh. 3 of 16, ARs A-C5, A-D8, D-C2, & D-C3 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
VIRGINIA POWER 1-EI--CB-21A ANNUNCIATOR 05 1-AR-A-C5 NORTH ANNA POWER STATION REV. 1 APPROVAL: ON FILE Effective Date:11/20/97 NIS IR HI FLUX ROD STOP 1.0 Probable Cause 1.1 Either intermediate range channel current equivalent greater than 30% power 1.2 Malfunction of nuclear instrumentation 2.0 Operator Action 2.1 Stop power escalation. 2.2 IF Annunciator 1PD2 (P-10 PERM PR>10% BLK NIS LP TRIPS) is LIT THEN do the following: 2.2.1 Block both trains of Intermediate Range Block and ensure annunciator 1P-E2 (NIS IR TRNS A & B BLKD) is LIT. 2.2.2 Block both trains of Power Range Block Lo STPT and ensure annunciator 1PD3 (PR LOSP TRNS A & B BLKD) is LIT. 2.3 IF 1P-D2 (Pb PERM PR>10% BLK NIS LP TRIPS) is NOT LIT THEN refer to 1AP--4.2 for failure of an intermediate range channel. 3.0 References 3.1 Westinghouse functional diagram 5655D33 sh 3, 4 3.2 Westinghouse drawing 2716351 sh 11 3.3 117 15FE7Q6 3.4 lAP-4.2, Malfunction of Nuclear Instrumentation (Intermediate Range) 3.5 DCP-8801 3.6 DCP 97-803, Intermediate Range NI Flux Trip Setpoint Change 4.0 Actuation 4.1 Either intermediate range instrument bistable NC-35E or NC-36E and Rsx relay in NIS drawers N35 or N36
VIRGINIA POWER 1-EI-CB-21A ANNUNCIATOR D8 l-AR-A-D8 NORTH ANNA POWER STATION REV. 2 APPROVAL: ON FILE Effective Date:08/26/09 Nominal: 103% power on MIS PR 1/4 Channels HI FLUX (see NOTE below) ROD STOP NOTE: IF NIs were adjusted downward below the previous 100% power gain setting using l-PT-24, Calorimetric Heat Balance (Hand Calculation), when Reactor power was less than 90% OR l-PT-24.l, Calorimetric Heat Balance (Computer Calculation), when Reactor power was less than 85%, THEN see Reactor Data Book for setpoint. 1.0 Probable Cause 1.1 Reactor overpower (excessive steam demand) 1.2 Failure or misadjustment of 1 power range instrument 1.3 Auto rod control system failure 1.4 RCS Temperature high above program 2.0 Operator Action 2.1 IF due to Rx overpower, THEN reduce power to lOO% (or Rx Power Limit) . Refer to l-AP-38, Excessive Load Increase, as required. 2.2 IF due to misalignment of a power range channel, THEN run a calorimetic and readjust instrument. 2.3 IF due to failed instrument or bistable, THEN refer to 1-AP-4.3, Malfunction of Nuclear Instrumentation (Power Range) 2.4 IF auto rod control has failed, THEN take manual control of rods. 2.5 IF RCS Tavg is high above program, THEN restore using rods or boration. 3.0 References 3.1 W 5655D33 SH 4, 9 3.2 ll7ls-ESK-1OA 3.3 Westinghouse NIS tech manual 3.4 l-AP-4.3, Malfunction of Nuclear Instrumentation (Power Range) 3.5 l-AP-38, Excessive Load Increase 3.6 CTS 02-98-2172-001, Revise calorimetric heat balance procedures as stated in ET NAF-980068 4.0 Actuation 4.1 One out of four power ranges, greater than 103% (or setpoint) of full power, bistables NC 4lL, 42L, 43L, 44L in power range drawer A
- END -
VIRGINIA POWER l-EI-CB-21D ANNUNCIATOR C3 l-AR-D-C3 NORTH ANNA POWER STATION REV. 2 SNSOC APPROVAL: ON FILE Effective Date:05/06/02 NIS INTER Current equivalent to RGE HI FLUX 35% power on RX TRIP 1/2 Channels Interlocked with P-l0 1.0 Probable Cause 1.1 Uncontrolled RCCA bank withdrawal from a subcritical position 1.2 Failure to block permissive interlock PlO 1.3 Intermediate Range NI failure 2.0 Operator Action 2.1 IF the Reactor is tripped, THEN GO TO 1-E0, Reactor Trip or Safety Injection. 2.2 IF the Reactor is NOT tripped, THEN verify Reactor power level is normal by using alternate and redundant indications. 2,3 IF Reactor power level is NOT normal OR unable to determine that the Reactor is in a safe operating condition, THEN trip the Reactor and GO TO lE0, Reactor Trip or Safety Injection. 2.4 IF Reactor power level is < 10% AND at least 1/2 NIS IR RX TRIP CHNL I / II (Panel L-Bl, B2) status alarms are LIT, THEN begin an orderly shutdown and initiate Actions in accordance with Tech Spec 3.0.3 (ITS 3.0.3) for inoperable Solid State Protection System. 2.5 IF Reactor power level is > 10% but < 35% AND the logic of Step 4.1 exists, THEN begin an orderly shutdown and initiate Actions in accordance with Tech Spec 3.0.3 (ITS 3.0.3) for inoperable Solid State Protection System. 2.6 IF Reactor power level is > 35% AND PiC PERM PR > 10% BLK NIS LP TRIPS (Panel P-D2) is NOT LIT, THEN begin an orderly shutdown and initiate Actions in accordance with Tech Spec 3.0.3 (ITS 3.0.3) for inoperable Solid State Protection System. 2.7 IF all plant parameters are normal AND the alarm is due to unknown instrument malfunctions, THEN immediately contact Instrument Department to determine cause of alarm. 3.0 References
3,1 Precautions, Limitations, and Setpoints Document 3.2 Westinghouse Logic NA-DW-5655D33 3.3 11715ESK1OD, 1OAAA ( 3.4 Tech Spec 3.0.3 (ITS 3.0.3) 3.5 1E0, Reactor Trip or Safety Injection 3.6 DCP 97803, Intermediate Range NI Flux Trip Setpoint Change 4.0 Actuation 4.1 Reactor shutdown logic: EITHER of the following logic conditions satisfied:
- Pb PERM PR > 10% BLK NIS LP TRIPS (Panel PD2) is NOT LIT
- at least 1/2 NIS IR RX TRIP CHNL I / II (Panel LB1, B2) status alarms is LIT 4.2 1NINC35F 4.3 1NINC36F
-END-
VIRGINIA POWER 1-EI-CB-21D ANNUNCIATOR C2 1-AR--D-C2 NORTH ANNA POWER STATION REV. 1 SNSOC APPROVAL: ON FILE Effective Date:05/06/02 NIS PWR RGE 25% power on HI ø-LO SP 2/4 Channels RX TRIP Interlocked with PjO N 1.0 Probable Cause 1.1 Uncontrolled RCCA bank withdrawal from a subcritical condition 1.2 Failure to block Power Range low setpoint Reactor Trip 2.0 Operator Action 2.1 IF the Reactor is tripped, THEN GO TO l-E0, Reactor Trip or Safety Injection. 2.2 IF the Reactor is NOT tripped, THEN verify Reactor power level is normal by using alternate and redundant indications. 2.3 IF Reactor power level is NOT normal OR unable to determine that the Reactor is in a safe operating condition, THEN trip the Reactor and GO TO l-E0, Reactor Trip or Safety Injection. 2.4 IF Reactor power level is < 10% AND at least 2/4 NIS PR LOSP RX TRIP CHNL I / II / III / IV (Panel L-C1, C2, C3, C4) status alarms are LIT, THEN begin an orderly shutdown and initiate Actions in accordance with Tech Spec 3.0.3 (ITS 3.0.3) for inoperable Solid State Protection System. 2.5 IF Reactor power level is > 10% but < 25% AND the logic in Step 4.1 exists, THEN begin an orderly shutdown and initiate Actions in accordance with Tech Spec 3.0.3 (ITS 3.0.3) for inoperable Solid State Protection System. 2.6 IF Reactor power level is > 25% AND Pl0 PERM PR > 10% BLK NIS LP TRIPS (Panel PD2) is NOT LIT, THEN begin an orderly shutdown and initiate Actions in accordance with Tech Spec 3.0.3 (ITS 3.0.3) for inoperable Solid State Protection System. 2.7 IF all plant parameters are normal AND the alarm is due to unknown instrument malfunctions, THEN immediately contact Instrument Department to determine cause of alarm. References 3.1 Precautions, Limitations, and Setpoints Document 3.2 Westinghouse Logic NADW5655D33
3.3 117l5ESK--1OD, 1OAAA 3.4 Tech Spec 3.0.3 (ITS 3.0.3) 3.5 1E0, Reactor Trip or Safety Injection Actuation 4.1 Reactor shutdown logic: EITHER of the following logic conditions satisfied:
- Pb PERM PR > 10% BLK NIS LP TRIPS (Panel P-1D2) is NOT LIT
- at least 2/4 NIS PR LOSP RX TRIP CHNL I / II / III / IV (Panel LC1, C2, C3, C4) status alarms are LIT 4.2 lNINC41P 4.3 1NINC42P 4.4 1NINC43P 4.5 1NINC44P
-END-
C) C C SOURCE RANGE RENCTRR TRIT TOAER RANGE REACTOR TRIP INTERHEOIATE RANGE REACTOR TRIP POWER RANRE HIGH NEUTRON PLOT RATE REACTOO TRIP () 0- 0-SOURCE RANGE ALOOF CONTROL NOTES 1 2 1/NA1E I/NA2E I/N43C I/N44E br i MAIOGAL RESET INRTE RI 1 I MANUAL RERER INOTE 01 I NRNORL RESET (NOTE SO I MANUAL RESET INTIE Al (SHEET 4) ri ri II R r-1 PIØ bERET 4) III IA IA OTHER (NOTE RI II LOGIC TRAIN A 0 TRIP TRIP BYPASS - RUPRUS I/N 3)0 IJN S2A Nb RACK) INIO RACE) OEEAORGIZE 2/ SOURCE RANGE HR. TROTH A E 0 LOGIC TRAIN NEAT) A AlOE NEUTROR TLUT MANUAL TRW R1GR NEUTRON (HIGH SEOFCINTI IRA IN CONTROL A2ARTI FLOG RATE Y REACTOR TRIP REACTOR TRIP (SHEER 21 ir tRE, 1=1 I (SHEET 21 RT 1 RR flI H oolF ST ENERRIOE SOURCE RANGE A H. (EITHER LORIC TAR IN) OTHERS 1 TRRERS R C HIGH NEUTRON FLOG RESET REACON TRIP RITA NEUTRON FLUX REACTOR bRIE TUFtER 21 REACTCR TRIP REACTOR TRIP ISHEER 21 (SHEET 21 NDTEG
- 1. THE RTTLI(JANT MANUAL ULCCE CONTROLS CONUIST OF TAO CTNTRTLS TN THE CONTROL NUCLEAR STEAM SUPPLY SYSTEM ATROT TOT EACH RANUE. ONE FUR ETCH TRAIN. FUNCTIONAL DIAGRAMS
- 2. I/N 33A 5 IN LOSIC TRAIN 0. NUCLEAR INSTRUMENT E MAFIIJAL TRIP SIGNALS I/N T3R 0 IN LORIC TRAIN 0.
- 3. I/N 3HA 5 IN LOGIC TRAIN A. UNITS 1 g 2 I/N 3AA IS IN LOGIC TRAIN A.
- 4. I/N 4/A IS IN LOGIC TRAIN A.
I/N 410 IS IN LOGIC TRAIN B. S. TNT COMPUTER INPUTS ARE CONNECTEE TO TAlE CIRERIT, INOIRITOAL HON EACH TRAIN. A. MANUAL RESET CONTROLS CONSIST OF FOUR MOPENTART CONTROLS IN THE CONTROL RORM ONE CONTROL FOR EACH INSTRLPENR CHANNEL. ORIGINAL ISSUE
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QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 2. 002K4.07 002/BANK/NAPS/H/2/3.1/3.5/2/
Plant conditions are as follows:
- The crew is preparing to roll Unit I turbine.
- Reactor power is 8%.
- Turbine power is 0%.
- RCSTAVGis551°F.
Based on these conditions, PRZR program level is and letdown will automatically isolate if PRZR level decreases to A. 28.4%; 15% B. 28.4% ; 23.4% C. 32.7%; 15% D. 32.7% ; 27.7%
- a. Incorrect. First part incorrect but plausible since this is the no-load value for PRZR level; if the candidate erroneously believes the program is based on First Stage pressure vice Tave they will select this distractor. Second part is correct, unlike the deviation alarm function which is variable, the isolation alarm function is fixed at this value.
- b. Incorrect. First part incorrect but plausible as noted above; second part is also incorrect but plausible since the candidate may confuse functions and setpoints for deviation and isolation alarms.
- c. Correct. First part is correct the level program is 28.4% to 64.5% based on Tave of 547°F to 580.8°F, so 551°F 32.7%. The second part is also correct; as noted above the isolation function is fixed at this setpoint, unlike the alarm function which varies with program (actual PRZR level 5%
below program PRZR level).
- d. Incorrect. First part is correct as discussed in Distractor c; second part is incorrect but plausible since as previously discussed the candidate may confuse the different alarm setpoints.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Reactor Coolant System (RCS) Knowledge of RCS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) Contraction and expansion during heatup and cooldown Tier: 2 Group: 2 Technical
Reference:
ARs B-F8 & B-G7, 1-SC-5.23, Westinghouse PLS book, dwg 108D014 sh. 4 of 17 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
VIRGINIA POWER 1-EI-CB-21B ANNUNCIATOR F8 1-AR-B-F8 NORTH ANNA POWER STATION REV. 0 APPROVAL: ON FILE Effective Date:11/08/96 PRZ LO 5% of Span below LEVEL Level Program 1.0 Probable Cause 1.1 Secondary load increase 1.2 Pressurizer control system failure (1-RCLC1459G) 1.3 Level Makeup control valve failure (1CHFCV1122) 1.4 Manipulation of the PZR level recorder selector switch l/1LR-459. 2.0 Operation Action NOTE: Decreasing PZR level may cause a coincident decrease in PZR pressure 2.1 Confirm proper operation of reactor control system 2.2 Confirm proper operation of pressurizer level control system 2.3 Monitor the operation of 1CH--FCV-1122 2.4 IF a transient is in progress, THEN stop or slow the transient if possible References 3.1 11715-FM93B 3.2 westinghouse Logic 108D014 3.3 Unit 1 Loop Book, page RC-68 4.0 Actuation 4.1 lRCLC1459E PZR Level
VIRGINIA POWER 1-EI-CB-21B ANNUNCIATOR G7 l-AR-B-G7 NORTH ANNA POWER STATION REV. 0 APPROVAL: ON FILE Effective Date:11/08/96 PRZ LO LEV HTRS OFF - 15% of Level LETDWN ISOL span 1.0 Probable Cause 1.1 Malfunction of pressurizer level control system 1.2 Excessive primary system cooldown 1.3 Instrument failure 2.0 Operator Action 2.1 Verify letdown isolation and heater cutoff 2.2 Verify operation of pressurizer level control system and reestablish level 2.3 Monitor containment sump level, temperature, pressure and radiation levels 2.4 WHEN level is restored, THEN restore letdown and PRZR heaters to service. 2.5 IF a PRZR level channel has failed, THEN GO TO lAP3, Loss of Vital Instrumentation 3.0 References 3.1 Westinghouse Logic 5655D33 3.2 11715FM93B 3.3 Unit 1 Loop Book, page RC068 4.0 Actuation 4.1 lRCLC1459C- Pressurizer Level 4.2 1RCLCl460C- Pressurizer Level
3= V g U C Cl) Li CL) C%J C V 0 C) H - 01W V L r V L N h__ H zC z V LL)C C D V V a 0 F-Cl)
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- V V
- (> ii peeipu
B. Spray Valve Controllers (PC-444C, PC-444D) Proportional Gain in % Spray 4%/% Controller Output (1) Valve Setpoint where spray is initiated on compensated 55% Controller Output (1) pressure signal from PC-444A C. Variable Heater Controller Proportional Gain in % Heating -6.6%/% Controller Output (1) Power Per PSI Setpoint where proportional 35% Controller heating is full on signal from PC-444A D. Power Relief Valve (PCV-455C) 92.5% Controller Output operated on compensated pressure signal from PC-444A to PC-444B E. Backup Heaters Turned On, On 30% Controller Output Compensated Pressure Signal From PC-444A to PC-444F F. Power Relief Valves (PCV-456 2335 PSIG operated on actual pressure) (PC-445A)
- 4. PRESSURE LEVEL CONTROL A. Level Program as Function of Tavg (TM-459)
Low Limit For Tavg = 547°F Lower limit will vary linearly for full power Tavg between 580.8°F and 586.8°F. For Full Load Tavg = 586.8°F 2 1.4% of span For Full Load Tavg = 580.8°F 28.4% of span Upper limit for full power 64.5% of span Tavg between 580.8°F and 5 86.8°F B. Low-Low Level Heater Cutout 15% of level span (LC-459C, LC-460C) 35
C PRESSURIZER LEVEL ( 0000EET 100 0001EET lOT ED E00000EL SET II ET000REL SET III _e___i [1 L TOOTTOF - TO TORT GOOD E:00Eo::_z TOG EIATS.OD SE_EGO r 0 j F 00L-O 500 TO&E 000SF iS FY00 To LOTI -, FOGY rot-ic -. El- e GOGH TORITROE LiZ I TO LTtOSOZ SOFT BAOTIOOI 05600050 LLEOTS II p OF SOlE IDES H 1 0 TTESOLR1OET TIOT OTTER LEROL OOEE LEOEL ITI50005500 RODE 01000 LOFOAI TO DOUbTS- 010100GL L459 TO TS110OF T0001TL L-000
° Qo LLO4OO LLETEI TO 601TROL T1000GL E-4S0 r:i 5°H L I_
SEE 500050 I FOR iS-ES & LETEMI t $ PROCESS INSTRUMENTATION & CONTROL CG100DET TO RELAY L*500.OO TO RELAY LE-000ED TO REtRY LERS-ELTO PRESSURIZER LEVEL CONTROL & PROTECTION LOGIC FOG PTTSSIAIDFTT ROCOLE ,EOTETS P-flS FLED 100t0 IOO, LOLO 150EOTITT. tOLD OLOOG RLOEK OIVORN-1 010005 (OTTO RODE 00000 REF OWL OS- IOTOGIO HOST TOGS 00001 UNITS I & 2 TE 0001+0 0G. 0000 NUCLEAR ENGINEERING NANTN ANNA POWER STATION TOO
° I TOTIOW I TOG NA-OW-108O014 --
P5:0001 TO USING FOR 0001610 ACRE 650165 TRIO FDA 0005 P51011+01 O0-TOA2NAD 00,0
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 3. 003 K5 .02 003/MODIFIED//H/3/2.8/3 .2/4/
Given the following conditions:
- Unit 1 is starting up following a mid-cycle forced outage.
- The crew has commenced slowly raising power in preparations for rolling the turbine.
At approximately 4% power, B RCP trips. Which ONE of the following identifies the effect of the RCP trip on the Departure from Nucleate Boiling Ratio (DNBR), AND includes the procedurally required action the crew will take to mitigate this event? A. DNBR has decreased stop power increase and be in Mode 3 within 6 hours. B. DNBR has decreased ; trip the reactor. C. DNBR has increased; stop power increase and be in Mode 3 within 6 hours. D. DNBR has increased ; trip the reactor.
- a. Incorrect. First part is correct; since RCS flow decreased CHF is lower so the ratio, or margin, is less.
Second part is incorrect. TS LCO is not met, however plant procedures direct tripping the Reactor in all cases where less than 3 RCPs are operating. The candidate who does not have detailed procedural knowledge may erroneously assume that the trip requirement only exists in Mode 1 (particularly since the DNB TS is ONLY applicable in Mode 1), whereas in this case we are in Mode 2. It should also be noted that this is a resent change.
- b. Correct. First part is correct as noted above. Second part is also correct, the AR for the annunciator directs tripping the reactor.
- c. Incorrect. First part is incorrect but plausible; as discussed in distractor a if the candidate does not understand the relationship between the RCP and this parameter they will be guessing as to the effect. Second part incorrect but plausible as discussed in Distractor a.
- d. Incorrect. First part is incorrect as discussed in Distractor c. Second part is corrrect as discussed above.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Reactor Coolant Pump System (RCPS) Knowledge of the operational implications of the following concepts as they apply to the RCPS: (CFR: 41.5 / 45.7) Effects of RCP coastdown on RCS parameters Tier: 2 Group: 1 Technical
Reference:
AR C-H6, TS 3.3.1 bases Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info: the procedural requirement to trip the reactor for these plant conditions was incorporated within the year
VIRGINIA POWER 1-EI-CB-21C ANNUNCIATOR H6 1-AR C H 6 NORTH ANNA POWER STATION REV. 4 APPROVAL: ON FILE Effective Date: 05/27/09 RC LOOP lB LO FLOW 90% 1 of 3 CH I-Il-Ill 1.0 Probable Cause 1.1 Channel in test 1.2 Channel failure 1.3 Failure of RCP 1.4 Failure of Loop stop valve 1.5 Breaker 15B3 opens 2.0 Operator Action ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 All Reactor Coolant Pumps - GO TO l-E-0, Reactor Trip RUNNING Safety Injection. 2.2 Check 1-RC-FI-1424, 1425, and 1426 to determine the affected channel.
/
2.3 Check flow in other loops. r 2.4 Determine if an approved test procedure has been initiated. 2.5 Only one channel indicating Go to Step 2.6. abnormal - GO TO l-AP-3, Loss of Vital Instrumentation. 2.6 Reactor power > 30%: Go to Step 2.7.
- a. At least 2/3 flow instruments a. GO TO l-E-0, Reactor Trip indicating abnormal due to Safety Injection.
instrument failure: 1-RC-FI-1424, Loop B RCS Flow CH I l-RC-FI-1425, Loop B RCS Flow CH II l-RC-FI-1426, Loop B RCS Flow CH III
- b. Commence an orderly shutdown and initiate Actions in accordance with Tech Spec 3.0.3.
2.7 Reactor power between 10% and 30%: Go to Step 2.8.
- a. At least 2/3 loop flow a. IF 2/3 loop flow instruments instruments indicating on 2/3 loops indicate low abnormal on 2/3 loops due flow, THEN GO TO l-E-0, to instrument failure. Reactor Trip or Safety Inj ection, Otherwise GO TO 1-AP-3, Loss of Vital Instrumentation.
- b. Commence an orderly shutdown and initiate Actions in accordance with Tech Spec 3.0.3.
2.8 Reactor power less than 10%:
- a. At least 2/3 loop flow a. At least 2/3 loop flow instruments indicating instruments indicate abnormal due to instrument low flow - Enter the failure GO TO l-AP-3, Loss Action of Tech Spec of Vital Instrumentation. 3.4.4 or 3.4.5.
3.0 References 3.1 Westinghouse Logic NA-DW-5655D33, sheet 5 3.2 ll7ls-FM-93A, Reactor Coolant 3.3 Instrument Loops 11715-RC-078, 079, 080 3.4 Westinghouse Solid State Protection System Interconnection Diagram NA-DW-108D744, sheet 3 3.5 Precautions, Limitations, and Setpoints Document 3.6 11715-ESK-1OC, 1OAAQ, 5AJ 3.7 Westinghouse Process Instrumentation & Control NA-DW-6007D09, D25, D44 3.8 Tech Spec 3.0.3, 3.4.4 and 3.4.5 3.9 1-E-0, Reactor Trip or Safety Injection 3.10 1-AP-3, Loss of Vital Instrumentation 4.0 Actuation 4.1 Relay K 126 solid state protection from l-RC-FC-1424 (C1-241) 1-RC-FT-1424 4.2 Relay K 214 solid state protection from l-RC-FC-l425 (C2-241) 1-RC-FT-1425 4.3 Relay K 313 solid state protection from 1-RC-FC--1426 (C2-233) l-RC-FT-l426 4.4 Relay K 238 from aux relay contact of breaker 15B3
NUCLEAR DESIGN INFORMATION PORTAL RTS Instrumentation B 3.3.1 BASES APPLICABLE 9. Pressurizer Water LevelHigh (continued) SAFETY ANALYSES, LCO, interaction concerns. The level channels do not actuate and the safety valves, and the high pressure reactor trip is APPLICABILITY set below the safety valve setting. Therefore, with the slow rate of charging available, pressure overshoot due to level channel failure cannot cause the safety valve to lift before reactor high pressure trip. In MODE 1, when there is a potential for overfilling the pressurizer, the Pressurizer Water LevelHigh trip must be OPERABLE. This trip Function is automatically enabled on increasing power by the P-7 interlock. On decreasing power, this trip Function is automatically blocked below P-i. Below the P-7 setpoint, transients that could raise the pressurizer water level will be slow and the operator will have sufficient time to evaluate unit conditions and take corrective actions.
- 10. Reactor Coolant FlowLow The Reactor Coolant FlowLow trip Function ensures that protection is provided against violating the DNBR limit due to low flow in one or more RCS loops, while avoiding reactor trips due to normal variations in loop flow.
Above the P-i setpoint, the reactor trip on low flow in two or more RCS loops is automatically enabled. Above the P-8 setpoint, which is approximately 30% RTP, a loss of flow in any RCS loop will actuate a reactor trip. Each RCS loop has three flow detectors to monitor flow. The flow signals are not used for any control system input. The LCO requires three Reactor Coolant FlowLow channels per loop to be OPERABLE in MODE 1 above P-i. In MODE 1 above the P-8 setpoint, a loss of flow in one RCS loop could result in DNB conditions in the core because of the higher power level. In MODE 1 below the P-8 setpoint and above the P-i setpoint, a loss of flow in two or more loops is required to actuate a reactor trip because of the lower power level and the greater margin to the design limit DNBR. Below the P-i setpoint, all reactor trips on low flow are automatically blocked since there is insufficient heat production to generate DNB conditions. (P North Anna Units 1 and 2 B 3.3.1-21 Revision 0
NUCLEAR DESIGN INFORMATION PORTAL RTS Instrumentation B 3.3.1 BASES APPLICABLE 11. Reactor Coolant Pump (RCP) Breaker Position SAFETY ANALYSES, [CO. Both RCP Breaker Position trip Functions operate from and three pairs of auxiliary contacts, with one pair on each APPLICABILITY RCP breaker with one contact supplying each train. These (continued) Functions anticipate the Reactor Coolant FlowLow trips to avoid RCS heatup that would occur before the low flow trip actuates. The RCP Breaker Position (Single Loop) trip Function ensures that protection is provided against violating the DNBR limit due to a loss of flow in one RCS loop. The position of each RCP breaker is monitored. If one RCP breaker is open above the P-8 setpoint, a reactor trip is initiated. This trip Function will generate a reactor trip before the Reactor Coolant FlowLow (Single Loop) trip setpoint is reached. The LCO requires one RCP Breaker Position channel per RCP to be OPERABLE. One OPERABLE channel is sufficient for this trip Function because the RCS FlowLow trip alone provides sufficient protection of unit SLs for loss of flow events. The RCP Breaker Position trip serves only to anticipate the low flow trip, minimizing the thermal transient associated with loss of a pump. This Function measures only the discrete position (open or closed) of the RCP breaker. Therefore, the Function has no adjustable trip setpoint with which to associate an LSSS. In MODE 1 above the P-8 setpoint, when a loss of flow in any RCS loop could result in DNB conditions in the core, the RCP Breaker Position (Single Loop) trip must be OPERABLE. In MODE 1 below the P-8 setpoint, a loss of flow in two or more loops is required to actuate a reactor trip because of the lower power level and the greater margin to the design limit DNBR. The RCP Breaker Position (Two Loops) trip Function ensures that protection is provided against violating the DNBR limit due to a loss of flow in two or more RCS loops. The position of each RCP breaker is monitored. Above the P-i setpoint and below the P-8 setpoint, a loss of flow in two or more loops will initiate a reactor (conti nued) North Anna Units 1 and 2 B 3.3.1-22 Revision 0
NUCLEAR DESIGN INFORMATION PORTAL RTS Instrumentation B 3.3.1 BASES APPLICABLE 11. Reactor Coolant Pump (RCP) Breaker Position (continued) SAFETY ANALYSES, LCO. trip. This trip Function will generate a reactor trip and before the Reactor Coolant FlowLow (Two Loops) trip APPLICABILITY setpoint is reached. The LCO requires one RCP Breaker Position channel per RCP to be OPERABLE. One OPERABLE channel is sufficient for this Function because the RCS FlowLow trip alone provides sufficient protection of unit SLs for loss of flow events. The RCP Breaker Position trip serves only to anticipate the low flow trip, minimizing the thermal transient associated with loss of an RCP. This Function measures only the discrete position (open or closed) of the RCP breaker. Therefore, the Function has no adjustable trip setpoint with which to associate an LSSS. In MODE 1 above the P-7 setpoint and below the P-8 setpoint, the RCP Breaker Position (Two Loops) trip must be OPERABLE. Below the P-7 setpoint, all reactor trips on loss of flow are automatically blocked since no conceivable power distributions could occur that would cause a DNB concern at this low power level. Above the P-7 setpoint, the reactor trip on loss of flow in two RCS loops is automatically enabled. Above the P-S setpoint, a loss of flow in any one loop will actuate a reactor trip because of the higher power level and the reduced margin to the design limit DNBR.
- 12. Undervoltage Reactor Coolant Pumps The Undervoltage RCPs reactor trip Function ensures that protection is provided against violating the DNBR limit due to a loss of flow in two or more RCS loops. The voltage to each RCP bus is monitored. Above the P-7 setpoint, a loss of voltage detected on two or more RCP buses will initiate a reactor trip. This trip Function will generate a reactor trip before the Reactor Coolant FlowLow (Two Loops) trip setpoint is reached. Time delays are incorporated into the Undervoltage RCPs channels to prevent reactor trips due to momentary electrical power transients.
(conti nued) North Anna Units 1 and 2 B 3.3.1-23 Revision 0
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 4. 004G2.1.19 004/NEW//L/2/3.9/3.8/2/
The Auxiliary Building Operator reports that the running charging pump, 1-CH-P-1A, seems to be making some unusual noises. Which ONE of the following identifies the PCS parameter that can be trended by the OATCs to evaluate the status of 1-CH-P-1A? A. Cubicle Temperature. B. Inboard Pump Bearing Temperature. C. Gear Box Cooler Outlet Service Water Temperature. D. Lube Oil Cooler Service Water Outlet Flow.
- a. Incorrect. Plausible since the parameter is available but not on PCS.
- b. Correct. This parameter can be directly viewed and trended on PCS.
- c. Incorrect. Plausible since this parameter is available in the CR but not on PCS.
- d. Incorrect. Plausible since there is a PCS point but it is the combined loads on the header not the specific Charging pump lube oil cooler.
Chemical and Volume Control System Ability to use plant computers to evaluate system or component status. (CFR: 41.10 /45.12) Tier: 2 Group: 1 Technical
Reference:
PCS printouts Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
Min Jarms Grd,Icc Trenac Po9ts Srch Zoom L s 1 Pnnt Hdp
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RT CH SI-I IN RD CN GROUP DISPLAY FOR 1CHTE40*
** dynamic group **
UDa:e rate secods POINT ID QUAL VALUE UNITS DESCRIPTION SYN 3 T2050A GOOD 74.3399 DEGE CHARGING P14P 1A 1TR INNER BRG I 1-CH-TE-101A1 T2055A GOOD 73.6896 DEGF CHARGING PMP 1A MTR OUTER BRG I 1-CH-TE-101A2 T2056A GOOD 76.8800 DEGE CHARGING PMP lB MTR INNER BRG I 1-CH-TE-1O1B-1 T2057A GOOD 74.8251 DEGE CHARGING P1P lB HTR OUTER BRG I 1-CH-TE-1O1B-2 T2090A GOOD 113.0246 DEGE CHARGING P1P 1C MTR INNER BRG I 1-CH-TE-1O1C-1 T2094A GOOD 126.2033 DEGE CHARGING PHP 1C MTR OUTER BRG I 1-CH-TE-1O1C-2 Y4058A GOOD 101.5932 DEGE CHARGING PHP 1A PMP INBD BRG I 1-CHTE-102A-1 Y4059A GOOD 108.8122 DEGF CHARGING PHP 1A PMP OTBD BRG I 1-CH-IEIO2A-2 Y4060A GOOD 110.3240 DEGF CHARGING PHP 1A PHP THR BRG T 1-CH-TE-102A3 Y4061A GOOD 105.2122 DEGF CHARGING PMP lB PMP INBD BRG T 1-CH-TE-102B-1 Y4062A GOOD 107.6152 DEGF CHARGING PMP lB PMP OTBD BRG I 1-CH-TE4O2B-2 Y4063A GOOD 109.6996 DEGE CHARGING PMP lB PMP THR BRG T l-CH-TE-102B*3 Y4064A GOOD 123.0291 DEGF CHARGING PMP 1C PMP INBD BRG T 1-CH-TE-102CJ Y4065A GOOD 126.6121 DEGE CHARGING PHP 1C PMP OTBD BRG T 1-CH-TE-102C2 Y4066A GOOD 147.4239 DEGE CHARGING PMP 1C PMP THR BRG T 1-CH-TE-102C-3 (S
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QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 5. 004K6.15 005/NEW//H/4/2.8/3.1/2/
Unit 1 is at 100% power with the following charging pump configuration:
- 1-CH-P-1A is running.
- 1-CH-P-1B is in AUTO.
- 1-OH-P-iC is tagged out for breaker inspections.
The OATC notes the following indications:
- 1-CH-Pl-1121 Discharge Header Pressure erratic.
- 1-CH-Fl-1122, Charging Flow-erratic.
- 1-CH-P-1A, Motor Amps erratic.
Based on these plant conditions, which ONE of the following identifies the cause of these indications AND includes the actions taken per 1-AP-49, Loss of Normal Charging, to mitigate this event? A. Gas binding of 1-CH-P-1A; place i-CH-P-1B and 1-OH-P-lA in PTL, then vent and start 1-CH-P-1B. B. Gas binding of 1-OH-P-lA; start 1-OH-P-lB and stop 1-OH-P-lA, then vent 1-OH-P-lA. C. Relief valve stuck open on 1-CH-P-lA; stop 1-CH-P-1A and close 1-CH-P-1A discharge MOVs, then start 1-CH-P-1B. D. Relief valve stuck open on 1-OH-P-lA; start 1-CH-P-IB, then stop 1-CH-P-1A and close 1-OH-P-IA discharge MOVs.
- a. Correct. 1-AP-49 examines these parameters for indication of gas binding. If gas binding has occured then any pump that is subsequently started must first be vented.
- b. Incorrect. First part correct as noted above. Second part is incorrect but plausible; the candidate who does not have detailed procedural knowledge may assume that this course of action is correct since it would not require letdown to be isolated, nor would it interupt normal charging
- c. Incorrect. First part is incorrect but plausible since the candidate who lacks detailed understanding of the system could confuse indications of the malfunction. Second part is also incorrect but plausible as these actions are contained in 1-AP-49 but are there to address a stuck open check valve on a pump that was swapped or tripped.
- d. Correct. First part is incorrect as discussed in Distractor c; second part is also incorrect but plausible for reasons similar to those described in distractor b..
QUESTIONS REPORT for NAPS 2010 RD NRC Exam rev3 Chemical and Volume Control System Knowledge of the effect of a loss or malfunction on the following CVCS components: (CFR: 41.7 I 45.7) Reason for venting VCT and pump casings while filling: vents must connect to LRS Tier: 2 Group: 1 Technical
Reference:
1-AP-49 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
Donon NORTH ANNA POWER STATION ABNORMAL PROCEDURE NUMBER PROCEDURE TITLE REVISION 14 1 -AP-49 LOSS OF NORMAL CHARGING (WITH TWO ATTACHMENTS) PAGE 1 of 18 PURPOSE To provide the instructions to follow in the event of a Loss of Normal Charging Flow. ENTRY CONDITIONS This procedure is entered when the following conditions exist:
- 1 -CR-Fl-i 122 is off scale high OR low OR erratic, or
- 1 -CH-Pl-1 121 is off scale low OR below normal, or
- 1 -CH-Pl-i 121 is erratic, or
- Charging Pump amps are abnormal OR erratic, or
- Charging Pump is suspected of gas binding.
CONTilUOUS USE
NUMBER PROCEDURE TITLE REVISION 14 1 -AP-49 LOSS OF NORMAL CHARGING PAGE 2 of 18 ACTION! EXPECTED RESPONSE I I RESPONSE NOT OBTAINED 1._ CHECK CHARGING PUMP FOR GAS D GOTOStep3. BINDING: D
- Running Charging Pump suspected of gas binding AND
- One of the following conditions exist:
D
- 1-CH-PI-1121, Discharge Header pressure ERRATIC OR EJ
- 1-CH-FI-1122, Charging flow
- ERRATIC OR tk D
- Motor amps - ERRATIC
NUMBER PROCEDURE TITLE REVISION 14 1 -AP-49 LOSS OF NORMAL CHARGING PAGE 3 of 18 H_STEP [j ACTION/EXPECTEDRESPONSE I I RESPONSENOTOBTAINED 2._ SECURE CVCS SYSTEM: a) Isolate letdown by closing the following valves:
- 1) Letdown Orifice Isolation Valves:
D
- 1-CH-HCV-1200A C
- 1-CH-HCV-1200B C
- 1-CH-HCV-1200C
- 2) Letdown Isolation Valves:
C
- 1-CH-LCV-1460A C
- 1-CH-LCV-1460B C b) Place non-running Charging Pumps in PTL C c) Place running Charging Pump in PTL C d) Thoroughly vent the Charging pump to be placed in service using ATTACHMENT 2, VENTING CHARGING PUMPS, while continuing with Step 10
/9
NUMBER PROCEDURE TITLE REVISION 14 1 -AP-49 LOSS OF NORMAL CHARGING PAGE 4 of 18 ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED
- 3. VERIFY CHARGING PUMP Do the following:
MANIPULATIONS IN PROGRESS a) Isolate letdown by closing the following valves:
- 1) Letdown Orifice Isolation Valves:
D
- 1-CH-HCV-1200A D
- 1-CH-HCV-1200B D
- 1-CH-HCV-1200C
- 2) Letdown Isolation Valves:
D
- 1-CH-LCV-1460A D
- 1-CH-LCV-1460B
\LI
- 4. CLOSE DISCHARGE MOVs ON D b) GOTO Step 10.
Do the following: PREVIOUSLY RUNNING CHARGING PUMP a) Place non-running Charging Pumps in PTL. D b) Place running Charging Pump in PTL. D c) Manually close 1 -CH-FCV-1 122, Charging Flow Control Valve. D d) WHEN Charging Pump discharge pressure has decreased to minimum, THEN start the previously running Charging Pump. e) Manually restore charging flow using 1-CH-FCV-1 122, Charging Flow Control Valve.
NUMBER PROCEDURE TITLE REVISION 14 1 -AP-49 LOSS OF NORMAL CHARGING PAGE 5 of 18 -i-{ ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED
- 5. VERIFY RUNNING CHARGING Do the following:
PUMP NORMAL: D
- 1-CH-PI-1121, Discharge a) Isolate letdown by closing the following valves:
Header pressure NORMAL D
- 1 -CH-FI-1 122, Charging flow - 1) Letdown Orifice Isolation Valves:
NORMAL C
- 1-CH-HCV-1200A D
- Motoramps-STABLE C
- 1-CH-HCV-1200B C
- 1-CH-HCV-1200C
- 2) Letdown Isolation Valves:
C
- 1-CH-LCV-1460A C
- 1-CH-LCV-1460B C b) GOTO Step 10.
NUMBER PROCEDURE TITLE REVISION 14 1 -AP-49 LOSS OF NORMAL CHARGING PAGE 6 of 18 j_STEP j ACTION/EXPECTEDRESPONSE I RESPONSENQTOBTAINED I 6.._ CHECK LETDOWN - IN SERVICE Do the following: D a) Control 1 -CH-FCV-1 122 to establish at least 25 gpm Charging flow. D b) Put 1-CH-PCV-1145 in MANUAL and open to 100%. c) Open the following Letdown Isolation Valves: lI
- 1-CH-TV-1204A D
- 1-CH-TV-1204B D
- 1-CH-LCV-1460A
- 1-CH-LCV-1460B d) Open one of the following Letdown Orifice Isolation Valves:
D
- 1-CH-HCV-1200A OR D
- 1-CH-HCV-1200B OR I1
- 1-CH-HCV-1200C lJ e) Adjust 1-CH-PCV-1145 to establish 300 psig letdown pressure and put 1-CH-PCV-1145 in AUTO.
D f) IF auto PRZR level control is desired, THEN place 1-CH-FCV-1122 in AUTO. (STEP 6 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 14 1 -AP-49 LOSS OF NORMAL CHARGING PAGE 7 of 18 H_STEP ACTION/EXPECTEDRESPONSE I RESPONSENOTOBTAINED I
- 6. CHECK LETDOWN - IN SERVICE (Continued) g) Maintain PRZR pressure at 2235 psig by operating the following:
D
- PRZR Heaters D
- PRZR Spray Valves NOTE: 1-CH-P-1C, C CHARGING PUMP, has no Auto-start features.
7._ CHECK STANDBY CHARGING D Place standby Charging Pumps in PUMPS - IN AUTO-AFTER-STOP AUTO-AFTER-STOP as directed by SRO.
- 8. SUBMIT WORK REQUEST FOR FAILED CHARGING PUMP DISCHARGE CHECK VALVE
- 9. RETURN TO PROCEDURE AND STEP IN EFFECT 1O. VERIFY VCT LEVEL GREATER- E1 Increase VCT level to greater than 12%.
THAN 12% 5 D IF level cannot be immediately restored, THEN I GOTOStep14.
NUMBER PROCEDURE TITLE REVISION 14 1-AP-49 LOSS OF NORMAL CHARGING PAGE 8 of 18 j_STEP ACTION/EXPECTED RESPONSE RESPONSE NOTOBTAINED I NOTE: Charging Pump suction could automatically transfer to the RWST depending on Charging flow rate, initial VCT level, VCT makeup flow rate, and the time taken to restore letdown.
- 11. VERIFY CHARGING PUMP D Manually open valves.
SUCTION FROM VCT ISOLATION VALVES OPEN: EJ
- 1 -CH-MOV-1 11 5C E1 H valves cannot be opened, THEN GO TO Step 14.
D
- 1-CH-MOV-1115E
- 12. VERIFY VCT PRESSURE - D Increase VCT pressure using 1-OP-8.6, VOLUME GREATER THAN 15 PSIG CONTROL TANK OPERATIONS, while continuing with this procedure.
13i GOTOSTEP15 14._ AÜG1TCAARGING PUMP SUCTION TO RWST: a) Open Charging Pump Suction From RWST Isolation Valves: IJ
- 1-CH-MOV-1115B D
- 1-CH-MOV-1115D b) Close Charging Pump Suction From VCT Isolation Valves:
D
- 1-CH-MOV-1115C D
- 1-CH-MOV-1115E
NUMBER PROCEDURE TITLE REVISION 14 1 -AP-49 LOSS OF NORMAL CHARGING PAGE 9 of 18 j_STEP ACTION/EXPECTEDRESPONSE I I RESPONSENOTOBTAINED I 15._ VERIFY CHARGING PUMP D Manually open MOVs. SUCTION MOVs OPEN: D
- 1-CH-MOV-1267A D
- 1-CH-MOV-1267B 11
- 1-CH-MOV-1269A C
- 1-CH-MOV-1269B C
- 1-CH-MOV-1270A C
- 1-CH-MOV-1270B 16._ CHECK CHARGING PUMP DISCHARGE PAtH:
a) Charging Pump Discharge MOVs - C a) Manually open MOVs. OPEN: C
- 1-CH-MOV-1286A C
- 1-CH-MOV-1287A C
- 1-CH-MOV-1286B C
- 1-CH-MOV-1287B C
- 1-CH-MOV-1286C C
- 1-CH-MOV-1287C (STEP 16 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 14 1 -AP-49 LOSS OF NORMAL CHARGING PAGE 10 of 18 STEP ACTION/EXPECTED RESPONSE RESPONSE NOTOBTAINED I
- 16. CHECK CHARGING PUMP DISCHARGE PATH: (Continued) b) Charging Line Isolation Valves - D b) Manually open valves.
OPEN: D
- 1-CH-MOV-1289A D
- 1-CH-MOV-1289B D c) 1-CH-FCV-1 122 OUTPUT
- c) Do one of the following:
DEMAND INDICATED D
- Open 1-CH-FCV-1122.
OR D
- Locally throttle 1 -CH-289, 1 -CH-FCV-1 122 Bypass.
OR D
- Use 1-AP-20, OPERATION FROM THE AUXILIARY SHUTDOWN PANEL, to shift 1 -CH-FCV-1 122 to local control in the Auxiliary Shutdown Panel and control charging flow.
OR (STEP 16 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 14 1 -AP-49 LOSS OF NORMAL CHARGING PAGE 11 of 18 j_STEP j. ACTION/EXPECTEDRESPONSE I RESPONSENOTOBTAINED I
- 16. CHECK CHARGING PUMP DISCHARGE PATH: (Continued)
- IF Unit 1 is in Mode 5, 6 or Defueled, THEN I place Loop Fill in service, as follows:
D 1) Close 1-EP-CB-26A, Breaker 13, 1-RC-HCV-1556A, B, C, Loop Fill Hdr Isol I Valve.
- 2) Open at least one of the following Loop Fill Hdr Isol Valves:
D
- 1-RC-HCV-1556A I D
- 1-RC-HCV-1556B
- 1-RC-HCV-1556C C 3) Control Charging flow using 1 -CH-FCV-1 160, Loop Fill Flow Control.
(STEP 16 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 14 1 -AP-49 LOSS OF NORMAL CHARGING PAGE 12 of 18 ACTION/EXPECTED RESPONSE RESPONSE NOTOBTAINED _STEP_[1
- 16. CHECK CHARGING PUMP DISCHARGE PATH: (Continued)
D d) 1-CH-HCV-1310 OPEN
- d) Do either of the following:
D
- Open 1-CH-HCV-1310.
OR IF Unit 1 is in Mode 5, 6 or Defueled, THEN place Loop Fill in service, as follows: D 1) Close 1-EP-CB-26A, Breaker 13, 1-RC-HCV-1556A, B, C, Loop Fill Hdr Isol Valve.
- 2) Open at least one of the following Loop Fill Hdr Isol Valves:
D
- 1-RC-HCV-1556A D
- 1-RC-HCV-1556B D
- 1-RC-HCV-1556C D 3) Control Charging flow using 1 -CH-FCV-1 160, Loop Fill Flow Control.
NUMBER PROCEDURE TITLE REVISION 14 1 AP-49 LOSS OF NORMAL CHARGING PAGE 13 of 18 H_STEP fI ACTION/EXPECTEDRESPONSE I I RESPONSENOTOBTAINED 1 7. VERIFY CHARGING PARAMETERS - Do the following: NORMAL:
- 1-CH-Pl-1121, Discharge D a) Charging Pump is running, THEN adjust Header pressure GREATER
- Charging flow to 25 gpm.
THAN 2300 PSIG D b) IF Charging Pump is NOT running, THEN close 1CH-FCV:fl22, Charging Flow Control Valve. D c) GOTOStepl9. D
- Charging Pump Motor amps -
STABLE D
- 1-CH-Fl-1 122, Charging flow -
NORMAL OR LJ
- 1-CH-FI-1160, Loop fill Hdr flow
- NORMAL 4 1
- 18. GOTOSTEP21
NUMBER PROCEDURE TITLE REVISION 14 1 -AP-49 LOSS OF NORMAL CHARGING PAGE 14 of 18 j_STEP ACTION/EXPECTED RESPONSE RESPONSE NOTOBTAINED I IJ (J(rt
- 19. CHECK FOR PIPING RUPTURE: Isolate any leakage found and notify Health Physics Department.
a) Level trending - NORMAL: D *VCT C *RWST C b) Auxiliary Building Sump Level - NORMAL 1 C c) Send an Operator to locally verify piping systems are intact d) Check Radiation Monitors - NORMAL: f) C
- 1 -RM-RMS-1 54, Aux Bldg Control Area C
- 1-RM-RMS-156, Sample Room C
- 1 -GW-RI-1 78-3, Process Vent RM Particulate C
- 1-GW-Rl-178-1, Process Vent RM Noble Gas Normal C
- 1-VG-Rl-179-3, Vent Stack A Particulate C
- 1 -VG-RI-1 79-1, Vent Stack A Noble Gas Normal C
- 1-VG-RM-105, Multi-Sampler Particulate (Points No. 4 and 6)
C
- 1-VG-RM-106, Multi-Sampler Particulate (Points No. 4 and 6)
NUMBER PROCEDURE TITLE REVISION 14 1 -AP-49 LOSS OF NORMAL CHARGING PAGE 15 of 18 -_STEP ACTION/EXPECTED RESPONSE I RESPONSE NOTOBTAINED I CAUTION: IF any Charging Pump is suspected of Gas Binding, THEN any Charging Pump to be subsequently started MUST be thoroughly vented using ATTACHMENT 2, VENTING CHARGING PUMPS. 2O. VERIFY RUNNING CHARGING Do either of the following: PUMP NORMAL: D
- 1 -CH-PI-1 121, Discharge D
- IF the Standby Charging Pump is NOT gas Header pressure NORMAL
- bound, THEN start the Standby Charging Pump.
- Motor amps - STABLE ir D
- 1 -CH-FI-1 122, Charging flow - OR NORMAL
- Cross-tie charging using 0-AP-48, CHARGING J L 7 PUMPCROSSCONNECT.
D
- 1 -CH-FI-1 160, Loop fill Hdr flow D IF Charging Pumps cannot be controlled from the
- NORMAL Control Room, THEN evaluate using 1-AP-20, OPERATION FROM THE AUXILIARY SHUTDOWN PANEL, to locally operate Charging Pumps from the Auxiliary Shutdown Panel.
- 21. MAKE UP TO RCS USING 1-OP-8.1, Maintain RCS inventory by using one of the OPERATION OF THE CHEMICAL following:
AND VOLUME CONTROL SYSTEM:
- Align Normal Charging Header D
- Seal injection flowpath.
D
- Align Alternate Charging EJ
- Safety injection flowpath.
Header
NUMBER PROCEDURE TITLE REVISION 14 1 -AP-49 LOSS OF NORMAL CHARGING PAGE 16 of 18 _STEP ACTION/EXPECTEDRESPONSE I -1 RESPONSENOTOBTAINED
- 22. VERIFY SEAL INJECTION FLOW TO Adjust seal injection flow as follows:
RCPs -7 TO 10 GPM EACH a) Locally unlock and throttle seal injection
\f isolation valves:
C
- 1-CH-318, A RCP Seal Injection Isolation Valve C
- 1-CH-314, B RCP Seal Injection Isolation Valve C
- 1-CH-310, C RCP Seal Injection Isolation Valve C b) Throttle 1 -CH-HCV-1 186, RCPs Seal Water Flow Control, as required.
- 23. DETERMINE IF LETDOWN SHOULD Do the following:
PLACED IN SERVICE: C
- Normal Charging - IN SERVICE C a) Continue attempts to restore Normal Charging.
C
- VCT IN SERVICE
- C b) WHEN Normal Charging is restored with Charging Pump suction aligned to the VCT AND C
- PRZR level 20% OR PRZR level is 20% or greater, THEN perform GREATER Step 24.
NUMBER PROCEDURE TITLE REVISION 14 1 -AP-49 LOSS OF NORMAL CHARGING PAGE 17 of 18 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED H_STEP
- 24. PLACE LETDOWN IN SERVICE: D Place Excess Letdown in service using 1-OP-8.5, OPERATION OF EXCESS LETDOWN.
D a) Control 1-CH-FCV-1122to establish at least 25 gpm Charging flow. D b) Put 1-CH-PCV-1145 in MANUAL and open to 100%. c) Open the following Letdown Isolation Valves: D
- 1-CH-TV.-1204A C
- 1-CH-TV-1204B C
- 1-CH-LCV-1460A C
- 1-CH-LCV-1460B d) Open one of the following Letdown Orifice Isolation Valves:
C
- 1-CH-HCV-1200A OR C
- 1-CH-HCV-1200B OR C
- 1-CH-HCV-1200C C e) Adjust 1 -CH-PCV-1 145 to establish 300 psig letdown pressure and put 1-CH-PCV-1145 in AUTO.
(STEP 24 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 14 1 -AP-49 LOSS OF NORMAL CHARGING PAGE 18 of 18 j_STEP ACTION/EXPECTED RESPONSE I RESPONSE NOTOBTAINED
- 24. PLACE LETDOWN IN SERVICE:
(Continued) D f) IF auto PRZR level control is desired, THEN place 1 -CH-FCV-1 122 in AUTO. g) Maintain PRZR pressure at 2235 psig by operating the following: D
- PRZR Heaters
- PRZR Spray Valves 25._ VERIFY CHARGING PARAMETERS - Do the following:
NORMAL:
- 1-CH-PI-1121, Discharge D a) Continue attempts to establish RCS makeup Header pressure GREATER
- flow.
THAN 2300 PSIG D
- 1-CH-Fl-1 122, Charging flow - b) Notify Operations Manager on Call.
NORMAL lJ
- Charging Pump Motor amps - D c) WHEN RCS makeup flow is established, THEN STABLE GO TO Step 26.
- 26. SUBMITWORKREQUESTSAS REQUIRED 27.z RETURN TO PROCEDURE AND STEP IN EFFECT END
NUMBER ATTACHMENT TITLE ATTACHMENT 1 -AP-49 1 REFERENCES REVISION PAGE 14 lofi
- 11715-FM-95B, CHEMICAL AND VOLUME CONTROL SYSTEM
- 0-AP-48, CHARGING PUMP CROSS CONNECT
- 1-AP-20, OPERATION FROM THE AUXILIARY SHUTDOWN PANEL
- 1 -OP-8.1, OPERATION OF THE CHEMICAL AND VOLUME CONTROL SYSTEM
- 1-OP-8.5, OPERATION OF EXCESS LETDOWN
- 1-OP-86, VOLUME CONTROL TANK OPERATIONS
- DCP 95-226, Charging Pump Interlock Modification Unit 1 a DR N 98-0361, Bypassing Safety Injection Flow To Intact RCS Loops In Mode 1-4
- SOER 97-01, Potential Loss Of High Pressure Injection and Charging Capability From Gas Intrusion
- DCP 99-006, Process Vent and Ventilation System Radiation Monitor System Replacement
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-49 2 VENTING CHARGING PUMPS REVISION PAGE 14 lof2 CAUTION: IF any Charging Pump is suspected of Gas Binding, THEN any Charging Pump to be subsequently started MUST be thoroughly vented. NOTE: Venting should be directed to the Auxiliary Building Sump using appropriate hose, as required.
- i. Notify Health Physics that the Charging Pump system will be vented to the Auxiliary Building Sump AND monitoring assistance is required.
- 2. Throttle open 1 -CH-566, Charging Pumps Suction Header Vent Valve, until a steady stream of water issues from vent.
3.
/Close 1 -CH-566, Charging Pumps Suction Header Vent Valve.
- 4. Do the following to Vent individual pumps to be started:
i a) IF 1-CH-P-1A is to be started, THEN do the following:
- 1) Open 1 -CH-498, 1 A Charging Pump Suction P1 Test Conn Isol Valve, until a steady stream of water flows.
- 2) Close 1-CH-498, 1A Charging Pump Suction P1 Test Conn Isol Valve.
- 3) Fully open 1 -CH-651, 1 A Charging Pump Outboard Vent Valve, to vent the discharge seal housing until the seal is full.
- 4) Close 1-CH-651 when seal is full.
- 5) Fully open 1-CH-652, 1A Charging Pump Inboard Vent Valve, to vent the discharge seal housing until the seal is full.
- 6) Close 1-CH-652 when seal is full.
- 7) Notify the Unit SRO that venting of 1-CH-P-1A is complete and the pump can be started.
(STEP 4 CONTINUED ON NEXT PAGE)
NUMBER ATTACHMENT TITLE ATTACHMENT 1 -AP-49 2 VENTING CHARGING PUMPS REVISION PAGE 14 2of2 b) IF QCH P 1 B is to be started THEN do the following
- 1) Open 1-CH-500, lB Charging Pump Suction P1 Test Conn Isol Valve, untila steady stream of water flows.
- 2) Close 1 -CH-500, 1 B Charging Pump Suction P1 Test Conn Isol Valve.
- 3) Fully open 1 -CH-653, 1 B Charging Pump Outboard Vent Valve, to vent the discharge seal housing until the seal is full.
- 4) Close 1-CH-653 when seal is full.
- 5) Fully open 1-CH-654, lB Charging Pump Inboard Vent Valve, to vent the discharge seal housing until the seal is full.
- 6) Close 1 -CH654 when seal is full.
- 7) Notify the Unit SRO that venting of 1-CH-P-18 is complete and the pump can be started.
c) IF 1 -CH-P-1 C is to be started, THEN do the following: C -
- 1) Open 1-CH-502, 1C Charging Pump Suction P1 Test Conn lsol Valve, until a steady stream of water flows.
- 2) Close 1-CH-502, 1C Charging Pump Suction P1 Test Conn Isol Valve.
- 3) Fully open 1 -CH-655, 1 C Charging Pump Outboard Vent Valve, to vent the discharge seal housing until the seal is full.
- 4) Close 1-CH-655 when seal is full.
- 5) Fully open 1 -CH-656, 1 C Charging Pump Inboard Vent Valve, to vent the discharge seal housing until the seal is full.
- 6) Close 1 -CH-656 when seal is full.
- 7) Notify the Unit SRO that venting of 1-CH-P-1C is complete and the pump can be started.
- 5. IF additional venting is required, THEN perform 1-PT-i 4.5, Venting ECCS Lines.
- 6. Notify the Unit SRO that Charging Pump venting is complete.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 6. 005K6.03 006/BANKJILT-SQ: 719 1/H/4/2.5/2.6/2/
Given the following conditions:
- RCS temperature is 175°F.
- RCS pressure is 305 psig.
- The PRZR is solid.
- Charging flow control is in manual.
- RHR is in service, with the A RHR heat exchanger and 1-RH-P-lA in service.
- 1-RH-.FCV-1605, RHR H/X bypass, is in AUTO.
The A RHR heat exchanger suddenly develops a 150 gpm tube leak. Assuming NO operator actions are taken, which ONE of the following identifies the effect of this malfunction on RHR pump flow, and on RCS pressure? A. RHR pump flow remains the same; RCS pressure remains the same B. RHR pump flow remains the same; RCS pressure decreases C. RHR pump flow increases; RCS pressure remains the same D. RHR pump flow increases; RCS pressure decreases
- a. Incorrect. Incorrect but plausible since if the candidate has the incorrect mental model (confusing Header Flow with pump flow) they may select this distractor. Second part also incorrect but plausible since the candidate may assume that 1-CH-PCV-1 145 will throttle closed to maintain pressure; while this is true the maximum letdown flow is 120 gpm and the given leak size is larger, so pressure must continue to decerease even after PCV-1 145 is fully closed.
- b. Incorrect. First part incorrect but plausible as noted above; second part is correct as explained in distractor a.
- c. Incorrect. First part is correct as noted above the leak creates a third flowpath in the system; since 1-RH-FCV-1605 is set in auto to maintain a constant return flow to the loops the FCV will throttle open and total pump flow will have to increase; second part incorrect but plausible as discussed in Distractor a.
- d. Correct. First part is correct as discussed above; second part is also corrrect as explained in distractor a.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Residual Heat Removal System (RHRS) Knowledge of the effect of a loss or malfunction on the following will have on the RHRS: (CFR: 41.7 I 45.7) RHR heat exchanger Tier: 2 Group: 1 Technical
Reference:
DWG 11715-FM-O94Ashs. 1&2, loop dwg RH 004 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
J3 2 R5
ø Vbø- MJ-gILII 030 *A4 ONIRMAG 7 8 5 4 3 I I rATES I
INSIDE REACTOR I OUTSIDE REACTOR I. FAG NOTES. SEE SPEET I OF THIS DERRING. CONTAINMENT RESIDUAL HEAT REMOVAL EXCHANGER 1-RH-8-IA RESIDUAL HEAT REMOVAL EXCHANGER I-RH E-IB 31 r A1 :::zcsc tLc c (IZe F cy.
- 2-.
(,
\ Cc SEE 117I5FM079R. --
1 SEE 11715+H-OOHR, - ( SH. I, E-4 C S c R SEE 117I5-FP-0750. Sn 1.0-4 P_ 520 A:WN 80T6 SEE 1I715-FR-0790. SH. I. C-4 CAS O-.) 5 cL)PJS (i4i -(--o THIS CAD DOAWING SYCETS I 6 2 OF 2 SPEERSEDES THE MANUAL PLOW WIG RAVE OPERATING NURYEYS 010670MG. IITI5-FM-MSFW REIIV AGO 1I7I5-FR-094A REV-Il RESPECTIOELY. REFERENCE DYWAI065 s 1- -Fc-° NUMMER TIrLE YEZO IIE2OI 050 _D jE2M VELAN 8076 D p
- c *D V -
I. FOR ADDITIONAL REFERENCE DRAWINGS. SEE SHEET I OW THIS DRAWING.
- 02. 11715 FR080R
- 03. 1I715-PM-ORMD
- 04. I1715-FM-OMOC FLOW/ADA DIMGRAM.FUEL PIT CLNG & REF PURIP SYSTEM FLOW/VON OIBGRAM.SAMPL(NG SYSTEM FLOW/VON DIAORAM,VENT N DRAIN SYSTEM E
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- 06. 117l5-FM-OM5C
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- CL 602 I. 1715-CRM-094A. NH. 2 NI CLOSSIFICATION ROUIIORRY ORRAINC 4-
- 2. DELETED LC TO PRIMARY AGAIN 3. I1715-SPM-094R. SN. 2 ISI SYSTEM PRESSIPE TESTING AGRWTNG 00 PRIMARY AGAIN FLANGES - TRANSFER SYSTEM TRANSFER SYSTEM 112-06-84-602-03 [-...-j 11715-ER -OSOC. SN. 2)-
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VIRGINIA POWER NORTH CAROIJNA POWER NUCLEAR ENGINEERING SERVICES RICHMOND ,RIROINIA FLOW/VALVE OPERATING NUMBERS DIAGRAM RESIDUAL HEAT REMOVAL SYSTEM NORTH ANNA POWER STATION UNIT 1 VIRGINIA POWER MERISED PER lEER NMR-502S-C00 REVISED PER OCR MN-25N REVISED PER MP 0-MCM-0060-0I THIS DERRING SAGERSEDES RER 17 OsDir 05604 560W COO NO. C, \l]s\dgY\tl IOFIR2MDGBVV II THIS DRAWING SUPERSEDES YEN IS ORIGINAL THIS DRAWING SLPEMSEOES REV (0 ORIGINAL 20 FR IS DiP4N 004/ORG 01501 EliDE DEARIAG NO. REM. CRAIDRER L000ENOR 11715FM -0R4A 20
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- 04. 11015-FRI-ONTO FLOW/TON WI/G/TR.SWRFLINO S/STUN 1R/RT OS. IITIT-ER-ONER FLOT/TON TI000/R.TEST N U/TIN S/STEM TSR OW. 11015-EN-RN/C FLUT/TON TI000TO./OST 6 TOUTS SySTEM 4Th 00. 11015-ERI-OSIN FLOW/U/N /:TO/U.OONT U/ETCH 000:00 SOT/U SOT STSTEN SEE i1715-FN-05 B. S/S.
4 I-S-I-F: 0/. IITIS-FM-OSWW FLOW/TON /i/GS/.TE/CTN/ 000UIT SGTEM ON. :TTIS-ERI-ONW/ FLLU/TON O:UG/TO.NEAO TON CO00.N/IT S/STEM
- 10. /ITITEN00N/A FLOW//TN T:TG/U.EESIWJUL SLOT TENT/AL STSTCE W/W-\
II. IIOIS-FEIOSTW FLOW/UON T:TS/N/.OFCT B TOJTE OONVTSGL STSTEN I4
- 2. 0021 002 RET. 12.
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SEAL COOLER 3/4-OS/WOO-WA V SEAL COOLER I T/4-//-4S-OO2-/2 I 1-RH-E-2A 7 0/ JO/2.-NO-TN./O2.OW I/EN 0/4 3/40 LOT/ I-RH- C }- -V c P-IA I RESIOUAL HEAT No BW / HELUNRRS/ REMOVAL PUMP N/PRO/I/CT TNT/INST IRT42E02-U2 I 0-N/-s-Bee-NW I. IIOIS-CER-ONWT. S/I 1SI OLHSSIFICU/IN/I 0/0400/0/ 0/001/0 MOTOR
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- 3. I1015-SP/-ONNT. S/I ITT S/S/EN P/ESSU/E TEST/C /00/1100 T
J4 I W/4-N/-Nl-002-02 04 0/P//TOOT N//ES NOT AFPLICW/LC c 2 N N/O:OVL SOUL TOLL-//LE TELL-TALC C/GINO ORAlS 0051/0 TRAIN I I/////A/WJ I
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,.-I4W/F 2 N/CLEWR ENGINEERING SERVICES NICHMOND.VIRGINIA N/TO 2 PT , WS/ CL /02 CL 002 FLOW/VALVE OPERATING NUMBERS DIAGRAM I I4-NN-0-1T02-GI I TI /2 RESIDUAL HEAT REMOVAL SYSTEM A NORTH ANNA POWER STATION UNIT 1 VIRGINIA POWER I/C/lOOT PEN 4+ TEE 0-E/CTO-64E$-OT /0/ISE/ FOR N/Il NI-lOWE 6 NI-N/NW /00150/ PER U/TNT P/NO 0-1/00-0460-01 TERINOO,IE N/_S F/ON-) WE TO /PTS EN/RI/AT TOTS N/TWINS SLEETSECES /5/ 13 NCC.PICCA PER 5/V/ION RE/lEN FLOW & /N.WE LEON/N/TIE IN/ItEMS 510(010005 050/N OS/N SEPT C/NT NT. /4Ag/.lT5- I ROTS SNIN3 SLEERSECES /0/ IS THIS N/N/ISO SLPERSEN/W ROT 14 4+0405010 141 TIN/IT EN/l4+// 05 04+ ON/tIlED 0 16 IS 14 13 12 ELSW/TOLTO OPEN/STINT ICOtEMS OI-OV4 N/AN TOE/SEC TISOL E/CO WOIlC 5/. RET.
,N4101 .qG0LLOP . AT0LNH- 04 1
I
- CRESA/A LEOGENOT 11715FM -@R4A 16 7GLLoWi_JEo_
8 I I S t I 3 I 2 V LEVER) 1W-SOP- INNS 6:06 /Ws/dR//NI0N4UINUON4s I
AREA: R. C. FCV- 1605 COMPUTER RM. COMPUTER I/O LOGIC RM. ELEV: 0234 CONTINENTAL 7613 ELEV: CABINET 2 ELEV: 0272 COL: 5 FISHER ACTUATOR 472 1- E I CB -21 TAG 12BA54R FE- 1605 REF: FM-39A ORIFICE 1-EI-CB- 18C FAIL CLOSE FE-4BM 0-8500 6PM F0626A FK- 1A 12-RH- 15 P SYMBOL FK-IB TO RHR HEADER FE-36J 3-15 FE- 3SAA FE-4V PSIG 1-EI-CB- 123 RACK- 1-100 IT FK-IE FK-6A I ii I V REF:
-1 7778 I Y/BR I I I BR BW B W B W/B TB I IHANNX0OS c_ >
ICHS6NXB56 1 ____ 1 EHANNXBB7 . > B B W JB70 BW w I *,j,,1 JPNcf EIIA7 7 CD M. C. RM. 1-E1-CB-04 C) 12-10 1211 I22 1 RLVV-16C I ii-.i I 11-2 I ii- }RCPv-7E AREA: SERV. WTR. ELEV: 0277 VERT. BOARD TAJ 33 34 TX 32 33 uJ LJ LJ ELEV: 0252 SECT. 1-2 0 CDL: 5.2/DE REF: ESK-IBBAJ B W FE-3BE C) REF: 13075-FE-7DM FE-3EA ICHSSNXO55 FE-3B0 1- IRHSA0I 1- IRHSB0I I IHA6NXO400 ci FE-7M F 1-1605 1-RHR-P- 1A 1-RHR-P- lB C) II J 15 WEST. VX252 LL ICHSGNXO2 I 1AM 0-8500 6PM w NN
- 1-EI-CB-301C B C) 1MUX- 14A INST. RK. RM.
ELEV: 0252 19 120 21 frBN 4-1-E1-CB-56 J36 3 I2 I
+
I TBN 16 11 J1 Y/B RHR LU
- 4- BR -- SYSTEM CD REF: FE-4V I IHA6NX001 I IHC6NXO02 HE LU FLOW m Li NA-DW-60080 17 ISCISNXOO9 BENCH BD. 1-21 1-EI-CB-02 J 13 C)
INTERNALS ICHS6NXO4G I H,i FC- 1605C FQ- 1605 HAGAN CD FC- 1605B r 0 4 0 U REF: FE-3X L________ MODEL 662O4%03 Cn FCY- 1605B 7 LU oPrd/t tJ C) FC- 1605A PROJECT: FY/1605A VIRGINIA POWER C)
*1
*1 z
FO! 1605 FM- 1605 -ii: :Y/BCONDUCTOR £ RESIDUAL HEAT REMOVAL SYSTEM NAPS NO. I LL CD J22 JOB NO: Fboc RHR HEADER FLOW INDICATION E U, & (JX CiD / .. J &QS- -c-tboc +0 CONTROL AND ALARM 11715 C) LOOP ACTION: FCV- 1605 IS CONTROLLED BY FT-1605. AN INCREASE IN FLOW WILL CAUSE FT-1605 TO ope REVISION OSCRIPTION OSGN REV
& 0)
(I I C) l el CLOSE FCV- 1605 MAINTAINING CONSTANT FLOW. THERE IS ALSO A REMOTE MANUAL o rJ -r tt REVISED PER DCP 00-118/OCR 2001-1361 NUMBER: C) c p; (d cz THIS DW6 SUPERSEDES THE REV 8 ORIGIONAL ELC F- 1605 lx ci
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 7. 006K6.18 007/BANKI/H/3/3.6/3.9/3/
Unit 1 was at 100% power when a small-break LOCA occurred. The crew is currently performing 1-E-1, Loss of Reactor or Secondary Coolant. The Unit Supervisor is at step 6, Check if SI Can Be Terminated. The following conditions exist:
- Containment pressure is 22 psia and slowly decreasing.
- RCS pressure is 1085 psig and slowly increasing.
- PRZR level is 30% and slowly increasing.
- CETCs are 490°F and stable.
- RCS loop hot-leg temperatures are 475°F and stable.
- AFW has been throttled to 150 gpm to each SG.
- A ICCM indicates subcooling of 66°F.
- B ICCM indicates subcooling of 80°F.
Given these conditions, and SI termination criteria A. A ICCM is indicating correctly; are met B. A ICCM is indicating correctly; are NOT met C. B ICCM is indicating correctly; are met D. B ICCM is indicating correctly; are NOT met
- a. Incorrect. Saturation temperature for the given pressure is 556 so based on the given CETC temperature subcooling should indicate about 66. Second part is incorrect but plausible since the candidate may lack the detailed procedure knowledge of the different setpoints for normal and adverse CNTMT criteria, or may overlook the fact that they are in adverse CNTMT or believe that it is the higher CDA setpoint of 28 psia.
- b. Correct. First part is correct as noted above; second part is also correct since CNTMT pressure is
>20 psia the adverse number of 75 must be applied, therefore SI can not be terminated based on the current plant conditions.
- c. Incorrect. First part is incorrect but plausible if the candidate is either unable to manipulate steam tables sucessfully or does not understand that subcooling is a function of CETCs NOT Ths. Second part incorrect but plausible as discussed in Distractor a.
- d. Correct. First part is incorrect but plausible as discussed in Distractor c; second part is correct as discussed in distractor b.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Emergency Core Cooling System (ECCS) Knowledge of the effect of a loss or malfunction on the following will have on the ECCS: (CFR: 41.7 / 45.7) Subcooling margin indicators Tier: 2 Group: 1 Technical
Reference:
EOP E-1, steam tables Proposed references to be provided to applicants during examination: Steam Tables Learning Objective: Question History: additional info:
NUMBER PROCEDURE TITLE REVISION 24 1 -E-1 LOSS OF REACTOR OR SECONDARY COOLANT PAGE 6 of 26 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED H_STEP
- 6. CHECK IF SI CAN BE TERMINATED:
D a) RCS subcooling based on Core Exit TCs - D a) GOTO Step 7. GREATER THAN 25°F [75°F] b) Secondary heat sink: C b) GOTO Step 7. D
- Total AFW flow to intact SGs -
GREATER THAN 340 GPM OR D
- Narrow range level in at least one intact SG GREATER THAN 11%[22%] ( v-e(j- rceS D c) RCS pressure STABLE OR
- C c) GO TO Step 7.
INCREASING D d) PRZR level - GREATER THAN 21% [26%] d) Do the following: C
- Try to stabilize RCS pressure with normal PRZR spray.
C *GOTOStep7. C e) GO TO 1 -ES-i .1, SI TERMINATION, STEP 1. S
CONTINUOUS ACTION PAGE FOR 1-E-1
- 1. ADVERSE CONTAINMENT CRITERIA IF either of the following conditions exist, THEN use setpoints in brackets:
- 20 psia Containment pressure, OR D
- Containment radiation has reached or exceeded 1 .0E5 R/hr (70% on High Range Recorder).
- 2. RCP TRIP CRITERIA IF both conditions listed below exist, THEN trip all RCPs:
C
- Charging Pumps AT LEAST ONE RUNNING AND FLOWING TO RCS, AND C
- RCS subcooling based on Core Exit TCs LESS THAN 25°F [85°F].
- 3. CHARGING PUMP RECIRC PATH CRITERIA C
- IF RCS pressure decreases to less than 1275 psi9 [1475 psig] AND RCPs tripped, THEN close Charging Pump Recirc Valves.
C
- IF RCS pressure increases to 2000 psig, THEN open Charging Pump Recirc Valves.
- 4. SI REINITIATION CRITERIA IF either condition listed below occurs, THEN manually start Charging Pumps and align BIT:
C
- RCS subcooling based on Core Exit TCs LESS THAN 25°F [75°F], OR C
- PRZR level - CANNOT BE MAINTAINED GREATER THAN 21% [26%].
- 5. ECST LEVEL CRITERIA C WHEN the ECST level decreases to 40%, THEN initiate 1-AP-22.5, LOSS OF EMERGENCY CONDENSATE STORAGE TANK 1 -CN-TK-1.
- 6. SECONDARY INTEGRITY CRITERIA IF either of the following conditions exist AND the affected SG has NOT been isolated, THEN GO TO 1-E-2, FAULTED STEAM GENERATOR ISOLATION, STEP 1:
C
- Any SG pressure is decreasing in an uncontrolled manner, OR C
- Any SG has completely depressurized.
- 7. 1-E-3 TRANSITION CRITERIA IF either of the following conditions exist, THEN manually start Charging Pumps, align BIT, and GO TO 1-E-3, STEAM GENERATOR TUBE RUPTURE, STEP 1:
C
- Any SG level is increasing in an uncontrolled manner, OR C
- Any SG has abnormal radiation.
- 8. COLD LEG RECIRCULATION TRANSFER CRITERIA C H RWST level decreases to less than 23%, THEN GO TO 1-ES-i .3, TRANSFER TO COLD LEG RECIRCULATION, STEP 1.
- 9. QS TERMINATION CRITERIA C WHEN RWST level is less than 3% AND QS Pump amps are FLUCTUATING, THEN perform ATTACHMENT 3, TERMINATION OF QUENCH SPRAY.
- 10. CASING COOLING TANK LEVEL C WHEN the Casing Cooling Tank level decreases to 4%, THEN ensure CDA is reset, close i-RS-MOV-iOOA and 1-RS-MOV-iOOB and stop both Casing Cooling Pumps.
- 11. RCP CRITERIA C Seal injection flow should be maintained to all RCPs.
- 12. REACTIVITY CONTROL CRITERIA C An Operator should be sent to locally close and lock 1 -CH-2i7, PG to Blender Isolation Valve.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 8. 007EK2.02 008/BANKI/L/3/2.6/2.8/1/
Given the following conditions:
- Unit 1 is at 50% power.
- B and C MEW pumps are tagged out.
- A MEW pump trips.
The OATC attempted to trip the reactor, but the reactor trip breakers did NOT open. Assuming NO additional operator actions are taken, which ONE of the following identifies how the AMSAC System will function to mitigate this event? AMSAC will open the Rod Drive MG Set A. supply breakers immediately after level in 2/3 SGs reaches 13% narrow-range. B. supply breakers 27 seconds after level in 2/3 SGs reaches 13% narrow-range. C. output breakers immediately after level in 2/3 SGs reaches 13% narrow-range. D. output breakers 27 seconds after level in 2/3 SGs reaches 13% narrow-range.
- a. Incorrect. Plausible since candidate may lack detailed knowledge of the AMSAC system design and default to this distractor since the only basis for the time delay is to allow the RPS a chance to actuate; the fact that the b-b level RPS trip setpoint is 18% NR implies that the RPS system already had a chance to actuate.
- b. Correct. The breakers that AMSAC actuates and the associated time delay are correct.
- c. Incorrect. Breakers are incorrect (see discussion in distractor d), and the actuation has a delay associated with it as discussed in distractor a.
- d. Incorrect. Even if the candidate recalls the time delay feature they may not recall the specific breakers that amsac actuates; also since the MG set output breakers would interrupt power to the CRDMs without the delay associated from coastdown of the MG set due to the flywheel the candidate who lacks knowledge of the specific breakers actuated by AMSAC would tend to gravitate towards this distractor.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Reactor Trip Knowledge of the interrelations between a reactor trip and the following: (CFR 41.7 /45.7) Breakers, relays and disconnects Tier: 1 Group: 1 Technical
Reference:
1 1715-LSK-5-8J, 8L, 8K, 6N Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
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NUCLEAR CONTROL ROOM OPERATOR DEVELOPMENT PROGRAM NORTH ANNA POWER STATION MODULE NCRODP-77-NA REACTOR PROTECTION SYSTEMS This document is the property of, and contains information proprietary to, Dominion. It is to be used solely for the purpose of training company employees and is to be returned upon request. This document and such information are not to be reproduced, transmitted, disclosed or used otherwise, in whole or in part, without the prior written consent of the Director, Nuclear Training. Dominion Nuclear Training 10/08/07
Reactor Protection Systems Inputs to the AMSAC System the output relays, and to control room annunciator alarms and status lights. The outputs to safety-related circuits are wired through safety-related qualified Class I E isolation relays. The steam generator level signals are from the narrow range channels of each steam generator. (A steam generator CH 474, 475, 476; B steam generator CH 484, 485, 486; C steam generator CH 494, 495, 496). The turbine load signals are from turbine impulse pressure channels Ill and IV (P-446, 447). The AMSAC panel is located in the instrument rack room. The AMSAC panel is powered from the TSC uninterruptible power supply through a breaker panel located in the HP office in the TSC. AMSAC Actuation The AMSAC is initiated when the turbine load is greater than 38 percent and a Logic complete loss of FW is detected. Loss of FW is the condition of any 2 of the 3 level transmitters in any 2 out of 3 steam generators 13 percent of narrow range level span for greater than 27 seconds. The 27-second time delay is performed by the PLCs to allow the RPS to respond prior to actuating AMSAC. The time delay is set to ensure that the turbine is tripped within 30 seconds of the onset of the ATWS event. Turbine load is a newly installed permissive (C-20) for AMSAC and is initiated when 2 of 2 first stage turbine pressures indicate turbine power >38 percent. Permissive C-20 stays locked in when power goes above 38 percent and stays locked in for 6 minutes when power is reduced below 38 percent. The 27-second steam generator low level timer does not function unless C-20 for the respective PLC is present, and the local NORMAL-BYPASS switch is in the NORMAL position. Automatic When the AMSAC system is actuated, the following actions are initiated: Actuations Performed by
- Main turbine trip via 2OAST-1 and 2OAST-2.
AMSAC
- All 3 AFW pumps will receive start signals.
- The supply breakers for the rod drive MG set will receive a trip signal.
- The steam generator blowdown trip valves and sample isolation valves will receive a close signal.
Outputs from the AMSAC panel is through four AMSAC output relays powered from the AMSAC system. The output contacts of these 2 relays are installed across the relay control power to the component which is to be actuated. This configuration eliminates the use of slave relays. Two of the relays (3A1 and 3B1) are non-safety grade relays, and are energize to actuate. Their purpose is to trip the Rod Control MG set supply breakers. Each relay trips both MG supply breakers. If AMSAC actuates, and power is subsequently lost, these relays will reset to their normal state (not actuated). The remaining two relays (3A and 3B) are safety-grade relays, and also require power to actuate. Their purpose is to trip the turbine, start AFW pumps, and close the steam generator blowdown trip valves and surface NCRODP-77-NA Page 67 10/08/07
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 9. 007K4.01 009/NEW//H/3/2.6/2.9/5/
Unit 1 is at 100% power when annunciator lB-HI, PRZ RELIEF TK HI TEMP, is received. The OATC confirms the alarm is valid, and notes that PRT level, pressure, and temperature have been slowly increasing since assuming the watch. Which ONE of the following identifies the source of PRT inleakage that has caused these plant conditions AND includes the flowpath used to drain the PRT? A. Letdown Relief Valve 1-CH-RV-1203 leaking by; PRT is drained to the PDTT and the PDTT is pumped directly to the in-service Boron Recovery Tank. B. Letdown Relief Valve 1-CH-RV-1 203 leaking by; PRT is drained to the PDTT and the PDTT is pumped directly to the Gas Stripper. C. Reactor Vessel Flange 0-ring leakage; PRT is drained to the PDTT and the PDTT is pumped directly to the in-service Boron Recovery Tank. D. Reactor Vessel Flange 0-ring leakage; PRT is drained to the PDTT and the PDTT is pumped directly to the Gas Stripper.
- a. Incorrect. First part is correct, this is a source of hot water at higher pressure and from the choices provided is the one that would cause an increase in temperature and level. Second part is incorrect, but plausible if the candidate lacks detailed systems knowledge thay may conclude that this flowpath would be logical since Gas Strippers are generally associated with RCS letdown.
- b. Correct. First part is correct as noted above; second part is also correct as the larger volume PRT is drained to the smaller PDTT the PDTT will pump to the Gas Stripper first.
- c. Incorrect. First part is incorrect but plausible since it is a source of high temperature water, however this line goes to the PDTT; the candidate who lacks detailed systems knowledge may confuse PRT vs. PDTT. Second part incorrect but plausible as discussed in Distractor a.
- d. Incorrect. First part is incorrect/plausible as discussed in Distractor C; second part is correct as discussed in b.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Pressurizer Relief Tank/Quench Tank System (PRTS) Knowledge of PRTS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) Quench tank cooling Tier: 2 Group: 1 Technical
Reference:
ARs B-Hi, B-Gi, B-Fl, i-OP-5.7, dwg 11715-FM-090C Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
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VIRGINIA POWER l-EI-CB-21B ANNUNCIATOR Fl 1-AR-B-Fl NORTH ANNA POWER STATION REV. 0 APPROVAL: ON FILE Effective Date:11/08/96 PRZ RELIEF TK 14 psig HI PRESS J.%AJ 1.0 Probable Cause 1.1 Nitrogen regulator failure 1.2 High tank level or excessive leakage 1.3 Power operated relief or safety valve leaking or actuation 2.0 Operator Action Response Not Obtained 2.1 Check PRT pressure Submit WR. GREATER THAN 14 PSIG 2.2 Check RCS pressure LESS IF pressure greater than THAN REQUIRED PRZR PORV required setpoint, THEN SETPOINT determine cause. WHEN pressure decreases to LESS than required setpoint, THEN Go To Step 2.3. 2.3 Check PRZR PORVs - CLOSED Manually close PRZR PORV(s) IF any PORV can NOT be closed, THEN close its block NOV. 2.4 Check PG to PRT - CLOSED Manually close valve(s). 2.5 Check N2 to PRT - CLOSED Manually close valve(s). 2.6 Check PRT level - STABLE IF increasing, THEN GO TO l-AP-16, Increasing Primary Plant Leakage 2.7 Check PRT temperature Drain and refill PRT using NORMAL lOP5.7, Operation of the Pressurizer Relief Tank. 2.8 Vent PRT using 1-OP5.7, Operation of the Pressurizer Relief Tank 3.0 References 3.1 117l5ESK10BAJ 3.2 W Instrumentation and control diagrams 3.3 PLS Document 3.4 Unit 1 Loop Book, page RC41 Actuation 4.1 lRCPC1472
VIRGINIA POWER 1-OP-5.7 NORTH ANNA POWER STATION REVISION 9-Pi PAGE4OF 16 3.0 INITIAL CONDITIONS 3.1 Review the equipment status to verify station configuration supports the performance of this procedure. 4.0 PRECAUTIONS AND LIMITATIONS 4.1 Comply with the following guidelines when marking steps N/A:
- the conditional requirements of a step do not require the action to be performed, THEN mark the step N/A.
- IE any other step is marked N/A, THEN have the Shift Supervisor (or designee) approve and justify the N/A on the Procedure Cover Sheet.
4.2 PRT pressure should be maintained between 8 and 14 psig during normal operations. 4.3 PRT pressure should be maintained 3 psig while draining. 4.4 Unit 1 MUST NOT be in DEGAS mode during PRT draining to the PDTT. The PDTT discharge will be aligned to 1 -BR-EV-2A, Gas Stripper in the DEGAS mode and will cause an RCS dilution.
VIRGINIA POWER 1-OP-5.7 NORTH ANNA POWER STATION REVISION 9-Pi PAGE 5 OF 16 mit Verif 5.0 INSTRUCTIONS 5.1 Draining the PRT 5.1 1
. Verify Initial Condition is satisfied.
5.1.2 Review Precautions and Limitations. CAUTION: Unit 1 MUST NOT be in DEGAS mode during PRT draining to the PDTT. The PDTT discharge will be aligned to 1 -BR-EV-2A, Gas Stripper in the DEGAS mode and will cause an RCS dilution. 5.1.3 Verify Unit 1 is NOT in the DEGAS mode.
- 5. 1 .4 Verify a positive pressure is present in the PRT. [F NOT, THEN establish a positive pressure using Subsection 5.4 before continuing.
5.1.5 Open 1-RC-HCV-1523, PRZR RELIEF TANK DRAIN ISOL, to drain the PRT. 5.1.6 Monitor PDTT level on LI-DG- 101, PRIM DR TK LVL. 5.1.7 WHEN the desired level is obtained, THEN close 1-RC-HCV-1523, PRZR RELIEF TANK DRAIN ISOL. Completed: Date:
VIRGINIA POWER 1-OP-S .7 NORTH ANNA POWER STATION REVISION 9-Pi PAGE6OF 16 5.2 Filling the PRT 5.2.1 Verify Initial Condition is satisfied. 5.2.2 Review Precautions and Limitations. 5.2.3 Open l-RC-TV-1519A, CNTMT PG SUPPLY ISOL. NOTE: Filling the PRT will greatly increase PG header flow. The standby PG Pump should be available for auto-start. 5.2.4 Close 1-RC-HCV-1523, PRZR RELIEF TANK DRAIN ISOL. 5.2.5 Open 1-RC-HCV-1519B, PRZR RELIEF TANK MAKEUP WATER SUPPLY ISOL, to fill the PRT. I 5.2.6 WHEN the desired level is reached, THEN close 1-RC-HCV-1519B, PRZR RELIEF TANK MAKEUP WATER SUPPLY ISOL. 5.2.7 IF desired, THEN close 1-RC-TV-1519A, CNTMT PG SUPPLY ISOL. Completed: Date:
VIRGINIA POWER 1-EI-CB-21C ANNUNCIATOR C8 l-AR-CC8 NORTH ANNA POWER STATION REV. 2 APPROVAL: ON FILE Effective Date:03/29/00 PDTT PP DISCHG > 55 psig HI PRESS 1.0 Probable Cause 1.1 Hi level in gas stripper - lBR-TVripped 1.2 Failure of l-DGPCV103 (PDTT pump discharge PCV) 1.3 Improper valve position on discharge of PDTT pump 2.0 Operator Action 2.1 Check operation of gas stripper (1OP10.1). 2.2 Check operation of 1DGPCV-103, open if necessary. 2.3 Check system valve lineup. 3.0 References 3.1 11715-FM-90C and B, vent and drains 3.2 NAPS instrumentation DG003 3.3 11715ESK1OC, 10 BAG r 3.4 S&W switch setpoint document system sort pg. 26 3.5 1l715LSK343A 3.6 EWR 88215 3.7 DCP 00800, Setpoint Change for PDTT Pump Disch Alarm 4.0 Actuation 4.1 1DGPC100
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QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 10. 007K5 .02 01 0/NEW/1H13/3.1/3 .4/5/
Operators are performing 1-OP-il, Unit Startup from Mode 5 at Less Than 140°F to Mode 5 at Less Than 200°F, in order to establish a steam bubble in the PRZR. Which ONE of the following identifies the PRZR temperature that operators should expect the steam bubble to form as stated in 1-OP-il, AND includes the action to be taken once bubble formation begins? A. 425°F ; throttle closed 1-CH-PCV-1145, Letdown Pressure Control Valve. B. 425°F; throttle open 1-CH-PCV-1145, Letdown Pressure Control Valve. C. 450°F throttle closed 1-CH-PCV-1145, Letdown Pressure Control Valve. D. 450°F ; throttle open 1-CH-PCV-1 145, Letdown Pressure Control Valve.
- a. Incorrect. Temperature is correct per OP-li; action is incorrect but plausible since the candidate who is unsure of valve controller operations or CVCS system operation may only recall that the evolution discusses letdown flow. If they reason that letdown flow would increase and it is desired to maintain it less than the limit of 120 gpm for the demineralizer, they may erroneously conclude that this action is necessary.
- b. Correct. Temperature is correct; action is also correct--the procedure provides guidance for increasing letdown flow to aid in bubble formation.
- c. Incorrect. Temperature is incorrect but plausible since there are concerns for maintaining RCS pressure higher in order to ensure adequate seal leakoff during cold shutdown mode of operation, however at this point the procedure establishes more restrictive control bands for this evolution (drawing bubble). Second part is incorrect but plausible as discussed in distractor a.
- d. Incorrect. Temperature is incorrect but plausible as discussed in distractor c. Second part is correct as explained in answer b, but again the candidate must have a understanding of plant conditions established by the procedure for the evolution in order to conclusively determine that this is the correct response.
Pressurizer Relief Tank/Quench Tank System (PRTS) Knowledge of the operational implications of the following concepts as the apply to PRTS: (CFR: 41.5 / 45.7) Method of forming a steam bubble in the PZR Tier: 2 Group: 1 Technical
Reference:
1-OP-1.1 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info: Some controllers at NAPS are inverse acting; this leads to an additional source of confussion for candidates who do not have detailed knoledge of the various control systems.
Y% PROCEDURE NO: p DetihiThon 1-OP-ti REVISION NO: NORTH ANNA POWER STATION 56 PROCEDURE TYPE: UNIT NO: OPERATING PROCEDURE I PROCEDURE TITLE: UNIT STARTUP FROM MODE 5 AT LESS THAN 140°F TO MODE 5 AT LESS THAN 200°F SURV REACT REQ MGT REVISION
SUMMARY
- FrameMaker template 030.
- Incorporated OP 09-0118:
- Added Step 5.24 to add requirement to sample the pressurizer and the RCS for boron before the PZR heatup takes place.
- Incorporated DCP 07-156:
- Changed Bowser Filter System with Turbine Lube Oil Conditioner System in Attachment 1.
PROBLEMS ENCOUNTERED: j NO j YES Note: If YES, note problems in remarks. REMARKS (Use back for additional remarks.) SRO: DATE: CONTINUOUS USE
DOMINION 1-OP-1.1 North Anna Power Station Revision 56 Page 34 of 54 NOTE: The Administrative maximum heat up/cooldown limit for the PRZR is 90 degrees per hour. 5.26.6 Increase PRZR temperature to 425°F at 90°F/hr. 5.26.7 Every 30 minutes, record PRZR heatup on Attachment 2. 5.26.8 WHEN PRZR temperature reaches 425°F, THEN ensure that the RCS temperature is greater than 170°F and increase Letdown to greater than Charging. NOTE: Pressurizer Surge Line temperature could change due to insurge and outsurges. 5.26.9 Decrease PRZR level to 28 percent using 1-RC-LI-1462, PRZR Cold Cal Level Indicator, and Station Curve 1-SC-S .23.1, Pressurizer Level LI-462 (STARTUP) Actual Level (%) As A Function Of Pressurizer Temperature. NOTE: The Administrative Limit for the maximum zXT between spray water temperature and PRZR liquid temperature is 3 00°F. 5.26.10 Slowly establish Pressurizer Spray. IF RCS °2 is >0.1 ppm, THEN minimize spray flow to keep Hydrazine in the Pressurizer as long as possible. 5.27 WHEN wide range T is 195°F, THEN do the following: 5.27.1 Stabilize RCS T at 195°F. 5.27.2 Stabilize RCS pressure between 325 psig and 340 psig. 5.27.3 Align The Main Steam System for service using 1-OP-28.1, Operation of Main Steam System.
DOMINION 1 -OP-i .1 North Anna Power Station Revision 56 Page 40 of 54 (Page 2 of 8) Attachment 2 RCS/PRZR Heat-Up Curve 600 Cr) 0 400 a) U) U) G) 0 C) 0 ci) c3) C ci) 200 0 0 50 100 150 200 250 300 350 400 450 Cold Leg Temperature (Deg. F) Graphics No: PC935 (Unit 1)
QUESTIONS REPORT for NAPS 2010 RD NRC Exam rev3
- 11. 008A3.10 01 1/MODIFIED/NAPS/H/4/2.9/3 .0/8/
Both Units are at 100% power. 1-CC-PCV-11O, Component Cig Pps Recirc Hdr Pressure Control Vv, is 50% open and controlling CC system pressure. Due to a tagging error, the instrument air supply to 1 -CC-PCV-1 10 is inadvertently closed. When instrument air is isolated to 1-CC-PCV-110, indicated pressure on 1-CC-Pl-100, CC HX Outlet Pressure, will and 2-CC-PCV-210 will A. decrease; throttle closed. B. decrease; NOT change position. C. increase; throttle open. D. increase; NOT change position.
- a. Incorrect. Incorrect but plausible if the candidate does not have detailed systems knowledge they may assume the failure mode of the subject valve would be open in order to ensure minimum pump flow to preclude pump overheat, however this is not the case. The response of the opposite units valve goes hand-in-hand with a faulty assumption of the failure mode.
- b. Incorrect. First part incorrect but plausible as noted above; second part will also go hand-in-hand with the first part because the candidate who has only cursory systems knowledge would conclude that since 1 -CC-PCV-1 10 is controlling 2-CC-PCV-21 0 would already be closed.
- c. Correct. The failure mode of this valve on loss of air is closed so header pressure will increase as a consequence of isolating air to it. With the system in a normal alignment the opposite Unit valve will throttle open when header pressure increases by 5-10 psig.
- d. Incorrect. First part is correct as discussed in Distractor c; second part is incorrect but plausible since the candidate who lacks detailed systems knowledge may conclude that since 1-CC-PCV-110 is controlling that 2-CC-PCV-210 is not in the picture, or that the change in 1-CC-PCV-110 is not of great enough consequence to cause 2-CC-PCV-210 to reach a setpoint that would cause itto open.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Component Cooling Water System (CCWS) Ability to monitor automatic operation of the CCWS, including: (CFR: 41.7 /45.5) CCW pump instruments and their respective sensors, including flow, pressure, oil level, and discharge temperature Tier: 2 Group: 1 Technical
Reference:
1-op-Si. 1, dwg 1171 S-FM-079A Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
DOMINION 1-OP-51.1 North Anna Power Station Revision 28 Page 4 of 48 This procedure includes sections for aligning CC in a cross-connected lineup, and for shifting from cross-connected to split plant operation. When cross-connected, the CC recirc header PCV control valves (1-CC-PCV-i 10, 2-CC-PCV-210) should be set so that one is controlling pressure at 90-95 psig, about 50% demand open, and the other PCV is closed (it should be set 5-10 psig above actual CCHX outlet pressure). It should be remembered that, when the CC Systems are cross-connected, actions affecting CC on one Unit will also affect the other Units CC System. Component Cooling flow is throttled through the Seal Return Heat Exchanger to increase the Reactor Coolant Pump (RCP) seal injection temperature. The RCP #1 seal leakoff flow rates have been noted to decrease as CC temperature decreases. Westinghouse data indicates seal leakoff flow will increase as seal injection temperature increases.
2.0 REFERENCES
2.1 Source Documents 2.1.1 UFSAR 9.2.2, Component Cooling System 2.2 Technical Specifications 2.2.1 Tech Spec 3.7.19 2.2.2 TRM3.7.15 2.2.3 TRM Section 7.5 2.3 Technical References 2.3.1 Flow Diagram 11715-FM-79A, Component Cooling Water System 2.3.2 Flow Diagram 11715-FM-79B, Component Cooling Water System 2.3.3 Flow Diagram 117 15-FM-79C, Component Cooling Water System
Component Cooling System Major Control Valves Major Control Valves The following paragraphs describe the operation of the major control valves associated with the CC System. The major control valves are operated in the Control Room from either the vertical panels, safeguards panels, or the benchboard. Many of the components cooled by CC water have control valves located in the CC discharge line. These valves are controlled by signals generated in the system being cooled by CC water; therefore, the operation of these valves is discussed in the applicable system module, rather than in this module. Where similar components are used in the system, only the A component and/or Unit I component is described. Normally the surge tank is filled manually by opening the bypass valve around LCV-I00. A discussion of the control features associated with LCV-100 is presented in the following paragraph (see Figure 51-2-NA). The CC surge tank level control valve (LCV-100) is air operated but can also be operated manually. When surge tank level is low, instrument air opens the valve to allow makeup water from the Condensate System to enter the surge tank. The valve closes on high level. LCV-100 is set to maintain a level of approximately 70 inches from the bottom of the tank. The valve is located on the fourth floor of the Auxiliary Building, beneath the CC surge tank. A manual handwheel on the valve is used when the automatic makeup control circuitry fails. Normally the CC surge tank is filled manually via the bypass valve around LCV-100. Water hammer results when LCV-100 opens, due to the relatively high discharge pressure of the condensate pumps. The CC supply pressure is maintained at a constant value by pressure control valve (PCV-110), which is set to maintain approximately 90 to 95 psig at the discharge of the CC heat exchangers. As discharge pressure increases, PCV-1 10 opens to bypass more of the CC supply water back to the suction of the CC pumps. In effect, PCV-Il0 maintains system head loss (as seen by the CC pumps) constant and prevents runout of the CC pumps. The trip valves associated with the CC System are all pilot solenoid, air-actuated control valves, and are operated from the Control Room. The valves are designed to fail shut on loss of electrical power to the pilot solenoid or loss of instrument air pressure (with the exception of TV-I 07A!B and TV-I 08A/B). Each valve is provided with red (open) and green (close) position indicating lights. To open TV-IOINB (Combined RCP Thermal Barrier CC Return Ctmt Iso Trip Valves), TV-102A/B/C/D/E/F (RCP CC Return Ctmt Iso Trip Valves), TV-103A/B (RHR HX CC Return), or TV-I 04A/B/C (RCP CC Supply), all of the following conditions must exist:
- 1. electrical power available to the solenoid,
- 2. instrument air pressure available,
- 3. CLOSE pushbutton not depressed,
- 4. OPEN pushbutton depressed, and NCRODP-51-NA Page 28 05/16/07
I ID: 3872 Points: 1.00 Unit 1 was operating at 100% power with the following conditions.
- CC systems are split out
- Unit 1 is supplying common loads
- A loss of instrument air has occurred to component cooling backpressure regulating valve 1-CC-PCV-1 10 Based on the above, 1 -CC-PCV-1 10 will fail and CC flow to the RCPs will A. closed; increase B. open; increase C. closed; decrease D. open; decrease Answer: A Question I Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 0 Difficulty: 0.00 System ID: 24073 User-Defined ID: 3872 Cross Reference Number:
Topic: 3872: NO TOPIC Num Field 1: Num Field 2: Text Field: 008-A2.05 Comments: Associated objective(s): List the following information associated with component cooling water backpressure control valve PCV-1 10.
- Means available in the control room to determine abnormal heat exchanger outlet pressure
- Position to which the valve fails on a loss of instrument air VA NAPS OPS Page: 1 of 1 14 December 2009
I ID: 2445 Points: 100 Unit 2 is in the startup process following a refueling outage with the Component Cooling Systems in split-plant configuration during the outage. You have been instructed to cross-connect the unit-1 and unit-2 Component Cooling Systems. In order to prevent each units backpressure PCV from competing for control of system pressure, which of the following actions should be taken? A. The setpoint for one units PCV is adjusted lower than the setpoint for the other units PCV. B. One units PCV is manually isolated, and the other units PCV isadjusted, as required. C. One units PCV is failed open, and the other units PCV is adjusted, as required. D. The isolation valves for each units PCVs are manually throttled until the desired pressure is reached. Answer: A Question I Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for oractice? No Points: 1.00 Time to Complete: 0 Difficulty: 0.00 System ID: 26625 User-Defined ID: 2445 Cross Reference Number Topic: 2445: NO TOPIC Num Field 1: Num Field 2: Text Field: 008000 Comments: Associated objective(s): Explain how the setpoint for the component cooling water backpressure control valves are adjusted when the unit-i and unit-2 systems are cross-connected. VA NAPS OPS Page: 1 of I 14 December 2009
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 12. 008AK3 .02 0 12/MODIFIED//H13/3 .6/4.1/3/
Unit 1 is at 100% power. A PRZR safety valve has indications of leakage. The following conditions exist:
- PRZR pressure is 2220 psig and slowly increasing.
- PRT pressure is 10 psig and slowly increasing.
Which ONE of the following identifies the expected temperature indication downstream of the safety valve, AND includes the reason for the indication? A. Approximately 240°F because this is an isenthalpic process. B. Approximately 240°F because of the long distance between the safety valve and the temperature detector. C. Approximately 650°F because this is an isenthalpic process. D. Approximately 650°F because of the short distance between the safety valve and the temperature detector.
- a. Correct. This temperature corresponds to the saturation pressure for the PRT and is arrived at since this is a constant enthalpy process (i.e. minimal energy loss).
- b. Incorrect. Plausible because the candidate who is unsure of the process may recall the expectation that the temperature would be lower and erroneously attribute that to heat that is lost to ambient between the leaking valve and the associated temperature element.
- c. Incorrect. Plausible because it is a constant enthalpy process and if the candidate erroneously equates this to mean constant temperature process they will select this distractor since 650°F corresponds to the saturation temperature for the Pressurizer pressure.
- d. Incorrect. Plausible since the candidate who does not understand that temperature is significantly lower could justify this as a reason for the temperature indicator to read about the same as the Pressurizer vapor space.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open) Knowledge of the reasons for the following responses as they apply to the Pressurizer Vapor Space Accident: (CFR 41.5,41.10/45.6/45.13) Why PORV or code safety exit temperature is below RCS or PZR temperature Tier: 1 Group: 1 Technical
Reference:
NCRODP module, steam tables Proposed references to be provided to applicants during examination: Steam TablesThe temperature Learning Objective: Question History: additional info:
Pressure Control and Protection System Operational Considerations Relief Valve Operation and Indication. There are many indications available to the operator that a safety or PORV has lifted. Using the PORVs as an example, the possible indications are as follows:
- 1. PRT pressure, level and temperature indication,
- 2. PRESSURIZER RELIEF TANK HIGH PRESSURE alarm,
- 3. PRESSURIZER RELIEF TANK HIGH-LOW LEVEL alarm,
- 4. PRESSURIZER RELIEF TANK HIGH TEMPERATURE alarm,
- 5. Acoustical monitor indication,
- 6. PRESSURIZER SAFETY VALVE OR PORV OPEN alarm,
- 7. Tailpiece temperature indication, and
- 8. PRESSURIZER POWER RELIEF LINE HIGH TEMPERATURE alarm.
All the available indication should be used to determine if the PORV is open. No one indication is enough to determine the valve that has opened. For 5 example, the acoustical monitors are not immune to flow noise interference from other nearby noise sources. The PRT level could increase due to other relief valve operation (RHR System) or due to spray from the Primary Grade Water System. However, if PRT level, temperature, and pressure are increasing, and the PORV acoustically indicates open, these indications together provide adequate assurance that the PORV has lifted. If the PORV has lifted, other indications can be checked to verify the conclusion. The PORV relieves 650°F saturated steam to the PRT. Although one might expect the PORV tailpiece temperature indicator (TI-1463) to read at or near 650°F, in fact the reading should be very different. The steam undergoes an extremely rapid expansion that is almost isenthalpic (a constant enthalpy process). This means that the Btus of energy per pound mass remain constant. The expansion, however, causes the Btus to be contained in a much larger volume. This effect causes indication to read between 200°F and 300°F. These values are high enough to actuate the high temperature alarm, but are lower than many expect. Normal Operations Plant Startup. This section partially describes a plant startup, emphasizing the components discussed in this system description. It is not intended to be a complete treatment of the startup procedure. Operating procedure OP-I describes the plant startup from cold shutdown conditions (plant temperature less than 200°F) to critical reactor operations at 5 percent power. The following paragraphs assume that the RC System is initially less than 200°F with pressure being controlled by the operation of the RHR system. The RC System is filled and vented, a charging pump is started and an RCP is started. Pressure is maintained at about 325 psig. This pressure satisfies the pressure-temperature limits of the RC System as well as that of the RHR system. The heatup of the plant is accomplished by operating the RCP, adding heat due to pump operation, and energizing all pressurizer heaters. NCRODP-74-NA Page 53 01/15/08
QUESTIONS REPORT for NAPS 2010 RD NRC Exam rev3
- 13. 008G2.4.45 013/NEW//H/4/4.1/4.3/8/
Both Units are at 100% power. Component Cooling pumps 1-CC-P-lA and 2-CC-P-lA are running. The following Unit 1 alarms are received:
- 1G-B3, CC HX lA-lB CC OUTLET LO FLOW.
- 1C-C4, RCP lA-B-C THERM BARR CC HI/LO FLOW.
ALL other Unit 1 alarms are clear. Which ONE of the following explains the cause of these alarms, AND includes the appropriate operator response to address them? A. Breaker 15H12, Stub Bus Supply Breaker, inadvertently opened; manually start 1-CC-P-i B. B. Breaker 15H12, Stub Bus Supply Breaker, inadvertently opened; verify 1-CC-P-i B automatically starts after 20 seconds. C. 15H13, 1-CC-P-lA Supply Breaker, overcurrent trip; manually start 1-CC-P-i B. D. 15Hi3, 1-CC-P-lA Supply Breaker, overcurrent trip; verify 1-CC-P-i B automatically starts after 20 seconds.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- a. Correct. Opening of this breaker will cause loss of 1-CC-P-lA, and result in the alarms listed above.
Second part is also correct since the problem was not caused by a fault with the running pump there would be no automatic start signal generated; further, although the stub bus has no voltage, UV relaying is upstream of this breaker and is therefore unaffected, however the candidate who lacks detailed systems knowledge may not be aware of this.
- b. Incorrect. First part incorrect as noted above. Second part is incorrect but plausible since this pump has auto-starts associated with it; however, as noted above, if the loss of running pump is due solely to a loss of the stub bus, there is no auto start (BUS UV auto start is associated with H bus, vice coming directly from the stub bus feeding the pump, however the candidate who lacks detailed systems knowledge may not be aware of the difference).
- c. Incorrect. First part is incorrect, but plausible; the loss of the pump will result in the subject alarms, however, an additional CC pump trip alarm would also be received. The statement that all other Unit 1 alarms are clear makes this distractor incorrect, however the candidate who lacks detailed systems knowledge may be unsure of or overlook this fact. Starting the standby pump would not normally be necessary since this event would generate an auto-start of the standby pump, but based on the information provided the candidate may erroneously conclude that an additional failure has occurred requiring manual action.
- d. Incorrect. First part is incorrect as discussed above. Second part also incorrect but plausible since as discussed above if the candidate errroneously concludes that lack of voltage on the stub bus would generate a UV start signal for the opposite train pump they will select this distractor.
Component Cooling Water System (CCWS) Ability to prioritize and interpret the significance of each annunciator or alarm. (CFR: 41.10/43.5/45.3 / 45.12) Tier: 2 Group: 1 Technical
Reference:
ARs G-F5, G-E8, G-B3C-C4, G-C3, DWG 11715-ESK-5P Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
VIRGINIA POWER 1-EI-CB-21G ANNUNCIATOR F5 1-AR-G-F5 NORTH ANNA POWER STATION REV. 2 APPROVAL: ON FILE Effective Date:04-24-02 T COMP COOL PP 1A AUTO TRIP 1.0 Probable Cause 1.1 Undervoltage on IH Bus 1.2 Motorelectrical fault 2.0 Operator Action 2.1 Ensure standby pump starts. If required, start pump using l-OP-51.1 Component Cooling System. 2.2 Determine problem if possible. 2.3 Notify Shift Supervisor. 2.4 If required submit WR. 2.5 IF required, enter the Action Statement of Tech Spec 3.7.3.1 (ITS 3.7.19) OR 3.7.3.2 (ITS TRM3.7.15) for Unit 1 and Unit 2. References 3.1 11715-FM79A, Component Cooling 3.2 11715ESK5P 4.0 Actuation 4.1 Electrical fault relay 86 energized
VIRGINIA POWER 1-EI-CB--21G ANNUNCIATOR E8 1-AR-G-E8 NORTH ANNA POWER STATION REV. 2 APPROVAL: ON FILE Effective Date:0424-02 COMP COOL PP lB AUTO TRIP 1.0 Probable Cause 1.1 Undervoltage on 1J bus 1.2 Excessive pump flow 1.3 Motor electrical fault 2.0 Operator Action 2.1 Ensure standby pump starts. IF required, start pump using lOP51.l, Component Cooling System. 2.2 Determine problem if possible. 2.3 Notify Shift Supervisor. 2.4 If required submit WR. 2.5 IF required, enter the Action Statement of Tech Spec 3.7.3.1 (ITS 3.7.19) or 3.7.3.2 (ITS TRM3.7.15) for Unit 1 and Unit 2. 3.0 References 3.1 1171579A, Component Cooling System 3.2 11715ESK5Q 4.0 Actuation 4.1 Electrical fault relay 86 energized (l5Jl3)
VIRGINIA POWER 1EI-CB-21G ANNUNCIATOR B3 1-AR-G-B3 NORTH ANNA POWER STATION REV. 1 APPROVAL: ON FILE Effective Date:11/29/00 CC HX lAlB CC OUTLET Total Flow LO FLOW < 5000 GPM 1.0 Probable Cause 1.1 Line rupture 1.2 Loss of Component Cooling system 1.3 Containment trip valve closed 1.4 CC Hx isolated 2.0 Operator Action 2.1 Verify Component Cooling water flow to components. 2.2 Monitor containment sump level. 2.3 Monitor Component Cooling water surge tank level. 2.4 Verify major valve positions. 2.5 IF Component Cooling flow is lost, THEN GO TO 1AP-15, Loss of Component Cooling. (. 0 References 3.1 UFSAR chapter 9.2.2 3.2 11715-FM-79A, Component Cooling 3.3 11715ESK5P, 5Q 3.4 Unit 1 Loop Book, page CC 063 4.0 Actuation 4.1 Component cooling Hx outlet 1A, or lB low flow (1CC--FTbOA) or (bCCFTbOB)
VIRGINIA POWER 1-EI-CB-2lC ANNUNCIATOR C4 l-AR-C-C4 NORTH ANNA POWER STATION REV. 1 APPROVAL: ON FILE Effective Date:09/14/05 RCP lA-B-C THERM BARR 59 gpm Hi CC ITI/LO FLOW 36 gpm Lo 1.0 Probable Cause 1.1 Reactor coolant to component cooling leak - Thermal barrier cooler tube failure 1.2 Component cooling system high pressure - flow imbalance 1.3 Loss of component cooling 2.0 Operator Action 2.1 For Hi flow 2.1.1 Check l-CC-FI-ll6A, 116B and ll6C to determine which RCP has CC Hi flow condition. 2.1.2 Determine if RCP Thermal Barrier Tube Rupture has occurred by monitoring for increased RCS Leakage, such as VCT level decrease OR charging/letdown imbalance. 2.1.3 Ensure l-CC-TV-ll6A, 116B or ll6C is closed on a Thermal Barrier tube rupture for affected RCP. 2.1.4 Manually isolate CC to affected RCP as soon as possible. 2.1.5 Maintain seal water injection flow. Do not allow pump to operate if lower radial bearing reaches 225°F. 2.2 For Lo flow 2.2.1 Check 1-CC-FI-ll6A, 116B and 116C to determine which RCP has CC lo flow condition. 2.2.2 IF Seal Injection AND Thermal Barrier CC Flow have both been lost, THEN GO TO l-AP-33.2, Loss of RCP Seal Cooling. 2.2.3 Ensure l-CC-TV-ll6A, ll6B or ll6C are open for the affected RCP. 2.2.4 Readjust CC flow as necessary. 2.2.5 IF caused by loss of component cooling, THEN refer to 1-AP-15. 3.0 References 3.1 117l5-LSK 25-1 3.2 l1715-ESK 6MC, 1CC, 1OAAJ 3.3 l1715-FM-79B component cooling 3.4 NAPS instrumentation CC 065, 064, 066 Actuation 4.1 l-CC-FT-116A, B, and C
VIRGINIA POWER 1-EI--CB-21G ANNUNCIATOR C3 1-AR-G-C3 NORTH ANNA POWER STATION REV. 0 APPROVAL: ON FILE Effective Date:092696 CC HX OUTLET 50 psig LO PRESS 1.0 Probable Cause 1.1 Loss of operating pump from overload or undervoltage on bus 1.2 Line rupture 1.3 Placing a second Component Cooling pump on line 2.0 Operator Action 2.1 Verify alarm and ensure standby pump starts. 2.2 Determine cause for alarm and correct if possible. 2.3 Notify Maintenance or Electrical Department of pump failure. 2.4 Do not lower CC pressure to alarm setpoint before starting second pump. 2.5 Carry out 1AP---15, Loss of CC System. 0 References 3.1 11715FM79A, Component Cooling 3.2 11715ESK--5P 3.3 Unit 1 Loop Book, page CC 059 4.0 Actuation 4.1 1CCPSL--100
VIRGINIA POWER 1-EI-CB-2lG ANNUNCIATOR C3 l-AR-G-C3 NORTH ANNA POWER STATION REV. 0 APPROVAL: ON FILE Effective Date:092696 CC HX OUTLET 50 psig LO PRESS 1.0 Probable Cause 1.1 Loss of operating pump from overload or undervoltage on bus 1.2 Line rupture 1.3 Placing a second Component Cooling pump on line 2.0 Operator Action 2.1 Verify alarm and ensure standby pump starts. 2.2 Determine cause for alarm and correct if possible. 2.3 Notify Maintenance or Electrical Department of pump failure. 2.4 Do not lower CC pressure to alarm setpoint before starting second pump. 2.5 Carry out 1AP--15, Loss of CC System. 0 References 3.1 1l7l5FM79A, Component Cooling 3.2 117l5ESK--5P 3.3 Unit 1 Loop Book, page CC 059 4.0 Actuation 4.1 lCCPSL--lOO
I 4 I 3 2 I 1 I D D MUX INPUT SPARe 1N09053._______ S SPARE ZR _) cm rE C DC COMPOkNT CooLI.iC PuA Cl.T CCPAOI (ORNJ CABLE) I 2. _i j+/-_ a*rc n 5 4-
£L_H INTLK THIS r
ftLV. CKT 515 INTLK THIS CKT CKT 4 I TAH- .LTAH-Ai7 (C4Mo4 FOR C3 C4 ai cv T CCIOIA T CCIOB i-cc-r- i ) I,
- 3D -IENSH 08 17Z H 1 __c L__ 3 AR-LS-IA ANN 5I9
-H F- SPARE SPA.RE 3F_FENSHO8 ESK-IOBAF INSt IsFARE: S T J AR.LS-IA FSL-CC -O6AE c BREAKER 5iI3 bOA & bOB (. zi T85 Co.ir &C LSK 62X- 62Y-1- E 3A 14- FB 3W I 43- FA 3W B B INPUT -
MUX T1N09043 z I NOTE: I. PREFLX LL WIRES WITH 1CCPØI
- 2. FIELD TO REMOVE CNST CONTACTS FROM RELAY 2X- ES)<-5Q1
- 3. FROM REL.Y O-CESø5, ESK llI-i I Dominion NUCLEAR ENGINEERING RICHMOND.VIRGIN{A I ELEMENTARY DIGRAM-416øV CKTS I COMPONENT COOLING PUMP 1-CC-P-1t I NORTH NN POWER STATION - UNIT I A I
REVFSIDN DEscRiprioN CAD NO C\\dgr\NlESK5PHYBII I I REVISED & REDRAWN PER DR 2øø3-5øO I THIS DWG SUPERSEDES REV 15 ORIGINAL j ORWtNG 11715ESK-5P NO REV 16 I [ OSGN ELC jSCLE NONE UNLESS OTHERWtSE NOTED jsii OF I I 3 2 I 1 PCECI t MCI I ø5-HcY-23 l:44 PRIOR 10 USING FOR DESIGN WORK CHECK OMIS FOR WORK PENDING I COOPERE
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 14. 009EA1 .03 014/BANKJNAPS/L/2/3.2/3.2/3/
Given the following conditions:
- Unit 1 was initially at 100% power.
- An RCS LOCA occurred.
- The crew tripped the Unit and initiated safety injection.
Plant conditions degraded, and CDA automatically actuated several minutes later. Based on the above sequence of events, the RSHX SW Outlet Radiation Monitor Sample Pumps, 1-SW-P-5, 6, 7, and 8, started A. immediately after SI was initiated. B. 2 minutes after SI was initiated. C. immediately after CDA actuated. D. 2 minutes after CDA actuated.
- a. Incorrect. Plausible since the candidate who lacks detailed systems knowledge may not understand what signals start the pumps and place the system in service.
- b. Incorrect. Plausible as noted above; additionally there is a two minute start delay.
- c. Incorrect. CDA starts the subject pumps, but not immediately. A two minute time delay feature is provided based on the concern that the piping is completely filled (water solid) prior to the pump starting to ensure that the operation / indication of the monitor is not comprimised. The candidate who is not knowledgable of this important design feature may mis-diagnose system operation.
- d. Correct. CDA starts the subject pumps after a two minute time delay as explained above.
Small Break LOCA Ability to operate and monitor the following as they apply to a small break LOCA: (CFR 41.7 /45.5 /45.6) Low-pressure SWS activity monitor Tier: 1 Group: 1 Technical
Reference:
1 -E-0, 1 -GI P-3A, 1 -PT-66.3 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
CONTINUOUS ACTION PAGE FOR 1-E-0
- 1. ADVERSE CONTAINMENT CRITERIA IF either of the following conditions exist, THEN use setpoints in brackets:
D
- 20 psia Containment pressure, OR
- Containment radiation has reached or exceeded 1 .0E5 R/hr (70% on High Range Recorder).
- 2. SI FLOW CRITERIA
- IF SI is actuated AND High-Head Cold Leg SI flow is NOT indicated, THEN perform ATTACHMENT 6, MANUAL VERIFICATION OF SI FLOWPATH.
D
- IF SI is actuated AND no Low-Head SI Pump is running, THEN manually start available pumps.
- 3. RCP TRIP CRITERIA IF both conditions listed below exist, THEN trip all RCPs:
D* Charging Pumps AT LEAST ONE RUNNING AND FLOWING TO RCS, AND D
- RCS subcooling based on Core Exit TCs LESS THAN 25°F [85°F].
- 4. CHARGING PUMP RECIRC PATH CRITERIA
- IF RCS pressure decreases to less than 1275 psig [1475 psig] AND RCPs tripped, THEN close Charging Pump Recirc Valves.
D
- IF RCS pressure increases to 2000 psig, THEN open Charging Pump Recirc Valves.
- 5. ECST LEVEL CRITERIA D WHEN the ECST level decreases to 40%, THEN initiate 1-AP-22.5, LOSS OF EMERGENCY CONDENSATE STORAGE TANK 1 -CN-TK-1.
- 6. CDA ACTUATION CRITERIA IF Containment pressure exceeds 28 psia OR 20 psia with Containment Radiation greater than 2 R/hr, THEN do the following:
- a. Manually actuate CDA.
1J b. Ensure CC Pumps STOPPED. D c. Stop all RCPs. D d. Ensure QS Pumps RUNNING.
- e. Ensure QS Pump Discharge MOVs OPEN.
- f. Initiate the following Attachments, when directed by ATTACHMENT 4, EQUIPMENT VERIFICATION:
D
- ATTACHMENT 2, VERIFICATION OF PHASE B ISOLATION D
- ATTACHMENT 3, PRIMARY PLANT VENTILATION ALIGNMENT
- 7. FAULTED SG ISOLATION D SI is in progress, THEN ATTACHMENT 7, FAULTED SG ISOLATION may be used for guidance on faulted SG(s) isolation and AFW flow control.
- 8. RUPTURED SG ISOLATION D H SI is in progress, THEN ATTACHMENT 8, RUPTURED SG ISOLATION may be used for guidance on ruptured SG(s) isolation and AFW flow control.
- 9. CONTAINMENT RECIRC MODE CRITERIA C To prevent possible radioactive release from the RWST, VCT level should be maintained greater than 12%.
- 10. RCP CRITERIA C Seal injection flow should be maintained to all RCPs.
- 11. REACTIVITY CONTROL CRITERIA C An Operator should be sent to locally close and lock 1-CH-217, PG to Blender Isolation Valve.
NUMBER ATTACHMENT TITLE ATTACHMENT 1-E-O 2 VERIFICATION OF PHASE B ISOLATION REVISION PAGE 42 8of9 H_STEP_[j ACTION/EXPECTEDRESPONSE I RESPONSENOTOBTAINED I
- 6. VERIFY THE FOLLOWING AUTOMATIC OPERATIONS ON THE UNIT 1 VENTILATION PANEL: (Continued)
D d) Verify Iodine Filter Banks IN SERVICE FOR SAFEGUARDS VENTILATION: FILTER (RED) FILTER (RED) (H Train) (J Train) 2,3, & 4 2,3, & 4 NOTE: The sample pumps automatically start following a 2-minute time delay. The Low Flow Alarm is enabled after an additional 30 seconds.
- 7. VERIFY THE FOLLOWING SAMPLE PUMP D Notify SRO.
RED LOW FLOW LIGHTS NOT LIT ON THE UNIT 1 RADIATION MONITORING PANEL: NOT LIT NOT LIT NOT LIT NOT LIT 1-SW-P-8 .1-SW-P-6 1-SW-P-7
VIRGINIA POWER 1 -GIP-3A NORTH ANNA POWER STATION REVISION 2 PAGE 32 OF 43 ATTACHMENT 1 (Page 26 of 37) FIGURES 1-EI-CB-47E Unit 1, TIain A SSPS Output Cabinet ACTUiIION: Slave Relay K626 is actuated by Containment Isolation K626 Phase B Master Relay K506. CONTAINMENT ISOLATION PHASE B FUNCTIONS TB 616 1 9 1SWSNO9COI ______ 120 VAC
- U START Timer for 1-SW-P-5, Radiation Monitoring Sample Pump Pump starts in 10 ISWSNO9TDO -
FE-9FB 120 seconds ESK-6JP START Timer for 1-SW-P-8, 3.jf----I1J 1SWSN12CO1 120 VAC Radiation Monitoring Sample Pump Pump starts in 4I____f__Ii 1SWSN12TD0 120 seconds ESK-6JP FE-9EM TB 617 SPARE 41 CLOSE 1-CC-TV-lO5C, 1 CCPCO7CO 1 Recirc Air Cooler 1 -VS-E-2C 120 VAC 1CCPCO7COO __ Discharge ESK-6MF FE-3EN 51 CLOSE 1-CC-TV-lOlA, 1CCPAO3COI -* RCP Thermal Barrier BLOCKED 120 VAC 1CCPAO3CO4 (ESK-2C, Det. C) ESK-6MC 17 I CLOSE 1-CC-TV-102A, FE-4BG 1CCPAO4CO1 RCP C Return Header BLOCKED 120 VAC ICCPAO4CO3 (ESK-2C, Det. C) ESK-6MC FE-4BG CLOSE 1-CC-TV-102C, I CCPCO4CO 1 --*
- RCP B Return Header 120 VAC BLOCKED 1CCPCO4CO3 * (ESK-2C, Det. C)
ESK-6MC FE.-4BG Jumper (TB 613-li) Connections toA Safeguards Test Cabinet for resetting slave relay Jumper (TB 616-7) i2 test circuits Jumper (TB 835-9) W 724 1D72 (Sheets 3 and 10) NA-DW-1082H41, Sheet 27 NA-DW-108D744, Sheet 20 FE-4AK Graphics No. BP300 FIGURE 1-626
t PROCEDURE NO: DoninioRI I -PT-66.3 REVISION NO: NORTH ANNA POWER STATION PROCEDURE TYPE: UNIT NO: OPERATIONS PERIODIC TEST I PROCEDURE TITLE: CONTAINMENT DEPRESSURIZATION ACTUATION OPERATIONAL TEST TEST FREQUENCY: UNIT CONDITIONS REQUIRING TEST: 18 Months Modes 1,2, 3, and 4 SPECIAL CONDITIONS: None SURV ICCE REQ III PMT REVISION
SUMMARY
- FrameMaker template rev. 030.
- Incorporated OP 08-0626 and CA090907 by making the following changes: Added Reference Step 2.4.18.
Corrected procedure numbers in Step 3.16. Added new P&L 4.24 concerning columns on Attachment 1. Added Step 6.53. Changed Steps 6.91 and 6.94. Replaced original Step 6.93 with new Steps 6.96, 6.97, 6.98, and 6.99. Added additional columns on Attachment 1, CDA/ØB Pre-Test Valve Line Up List. Changed column titles throughout the procedure. Added Step 11 on Attachment 16. Added new Attachment 22, Discrepancy Log. REASON FOR TEST (CHECK APPROPRIATE BOX): Surveillance j Post-Maintenance Work Order Number: TEST PERFORMED BY (SIGNATURE): DATE STARTED: DATE COMPLETED: TEST RESULT (CHECK APPROPRIATE BOX): CONDITON REPORT NUMBER(S) AND DATE: Satisfactory Unsatisfactory Partial THE FOLLOWING PROBLEM(S) WERE ENCOUNTERED AND CORRECTIVE ACTIONS TAKEN: (Use back for additional remarks.) COGNIZANT SUPERVISOR or DESIGNEE: DATE: ADDITIONAL REVIEWS: DATE: CONTINUOUS USE
DOMINION 1-PT-66.3 North Anna Power Station Revision 45 Page 70 of 131 7.1.6 Time delays for various equipment are as follows:
- l-RS-P-1A 120 sec. +/- 12.0 sec.
- l-RS-P-1B 120 sec. +/- 12.0 sec.
- 1-QS-MOV-102A 300 sec. +/- 30 sec.
- l-QS-MOV-102B 300 sec. +/- 30 sec.
- l-RS-MOV-lOlA 55 sec. to 70 sec.
- 1-RS-MOV-1O1B 55 sec. to 70 sec.
7.1.7 The following pumps started on a CDA signal:
- 1-SW-P-5
- 1-SW-P-6
- 1-SW-P-7
- 1-SW-P-8
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 15. 01 0A3 .0.2 01 5/BANKJNAPS/H/3/3 .6/3.5/3/
Unit 1 is at 100% power. The OATC notes that all of the PRZR Backup Heater white status lights on the vertical board are ON, and PRZR pressure is 2260 psig and slowly increasing. Assuming no operator action is taken, which ONE of the following identifies the failed channel, AND includes the system response to this failure? A. 1-RC-PT-1444 failed ; the PRZR spray valves will modulate open. B. 1-RC-PT-1444 failed ; one PRZR PORV will open. C. 1-RC-PT-1445 failed ; the PRZR spray valves will modulate open. D. 1-RC-PT-1445 failed ; one PRZR PORV will open.
- a. Incorrect. Plausible since there are two control channels and the candidate who lacks detailed systems knowledge may confuse their functions. Second part is also plausible since this is the normal system response (i.e. if heaters were locked on for a downpower, this is how the system would respond) provided the controlling channel is functioning properly.
- b. Correct. Due to the malfunction, spray valves will not respond to the pressure increase, however the other channel will open the PORV when pressure rises to 2335 psig.
- c. Incorrect. As noted above the candidate who lacks detailed systems knowledge may confuse their functions; for this failure the spray valves would be unaffected and respond as described, however this failure would not result in pressure increasing as given in the stem (1445 failing high would cause heaters to energize, but pressure would be decreasing because 1445 failing high would also directly open a PORV).
- d. Incorrect. Plausible since as previously discussed the candidate may not fully understand the functions and operation of the channels, or be able to accurately analyze the various failure modes and effects.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Pressurizer Pressure Control System (PZR PCS) Ability to monitor automatic operation of the PZR PCS, including: (CFR: 41.7 / 45.5) PZR pressure Tier: 2 Group: 1 Technical
Reference:
ARs B-F7, B-E7, dwg 5655D33 sh. 11 of 16 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
VIRGINIA POWER 1-EI-CB-2lB ANNUNCIATOR F7 l-AR-B-F7 NORTH ANNA POWER STATION REV. 3 APPROVAL: ON FILE Effective Date:03/27/07 PRZ HI-LO 2310 psi HI Reset-2303psi PRESS 2215 Psi LO Reset-2223psi 1.0 Probable Cause L 0 1.1 Pressure control system malfunction I S 1.2 Load transient 1.3 Instrument failure 3S 1.4 Cooldown in progress 2.0 Operator Action 2.1 Verify alarm: 2.1.1 Check PZR pressure channels 2.1.2 IF pressure is normal, THEN submit WR 2.2 High Pressure 2.2.1 Verify proper operation of the pressure controller. IF necessary, THEN take manual control and return pressure to normal 2.2.2 Ensure all heaters are off 2.2.3 Ensure both spray valves are fully open 2.2.4 IF due to a plant transient, THEN stop or reduce the rate of the transient to allow pressure to return to normal 2.3 Low Pressure 2.3.1 IF controlled cooldown/depressurization is in progress, THEN return to procedure and step in effect. 2.3.2 Verify proper operation of the pressure controller. IF necessary, THEN take manual control and return pressure to normal. 2.3.3 Ensure all heaters are on 2.3.4 Ensure both spray valves and power operated reliefs are fully closed 2.3.5 IF due to a plant transient, THEN stop or reduce the rate of the transient to allow pressure to return to normal 2.3.6 Go to l-AP-44, LOSS OF REACTOR COOLANT SYSTEM PRESSURE 2.3.7 IF pressure dropped below 2205 psig, THEN refer to Tech Spec 3.4.1 2.4 IF alarm is due to a failure of l-RC-PT-1445, THEN refer to TS 3.4.11. 3.0 References 3.1 ll7l5-ESK-1OAAG 3.2 PLS Document 3.3 W Drawing 5655D33 Sheet 11 3.4 Unit 1 Loop Book, page RC-l08 3.5 EWR 92 - 049A 3.6 Tech Spec 3.4.1 and 3.4.11 3.7 ICP-P-l-P-445 Actuation 4.1 High Pressure-l-RC-PC-1445C 4.2 Low Pressure-l-RC-PC-l445B
VIRGINIA POWER 1-EI-CB-21B ANNUNCIATOR E7 1 -AR-B--E 7 NORTH ANNA POWER STATION REV. 0 APPROVAL: ON FILE Effective Date:11/08/96 PRESSURI ZER HI PRESS 2335 psi NOTE: PORV 1456 Auto-Opens coincident with this alarm. 1.0 Probable Cause 1.1 Extreme Load rejection 1.2 Instrument failure or testing f vo 2.0 Operator Action ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 CHECK RCS STATUS - OVER PRESSURE Close l-RCPCV1456. IF CONDITION 1RCPCV1456 cannot be closed, THEN close 1RC-MOV1535 Reduce RCS pressure as required
*Verjfy/open PORVs
*Verify/open PRZR Spray Vlvs
*place PRZR heaters in PTL 2.2 CHECK PRZR PRESSURE - STABLE AT IF PRZR pressure continues to OR TRENDING TO 2235 PSIG decrease, THEN GO TO 1-AP44, Loss of Reactor Coolant System Pressure.
2.3 Submit WR as applicable 3.0 References 3.1 117 15ESK1OAAH 3.2 PLS Document 3.3 W Drawing 5655D33 Sheet 11 3.4 Unit 1 Loop Book, page RC-70 4.0 Actuation 4.1 Pressure comparator 1RCPC1445AX
PRESSURIZER PRESSURIZER AUX. F.P. CHARONG LEVEL CHANNELS PRESSURE CHANNELS STATION PUMP MEDIAN/HI TAVG STATION FROM --1 PRESSURIZER MEDIAN SIGNAL LOW PRESSURE (2/3) SELECT CIRCUIT (SHEET 0) (SHEET 91 0 TRAIN B TRAIN A ADJUSTABLE
* -------- 0 TAVO z.
WITH CONTROLLER PP. S TAT ION I ADJUSTABLE LEVEL CHARNEL PRESSS-RE
)- REFERENCE SETPOINT WITHIN I I-SAl TCH (POSITION .2
- (P-F RET) CONTROLLER )L-L REF) NORMAL) Y
- SEL EC TEE) 2Z (NOTE 41 (NOTE 41 (P-P RET) S
?
(NOTE 4) (NOTE 4) (NOTE 4) ALL ORIFICE ISOLATION VALVES CLC S CO POWER RELIEF VALVE POWER RELIEF VAL\C CONTROL MODE CONTROL MODE SEECIOR SWITCH SELECTOR SWITCH (CONTROL BOARD) ICONTROL BOARD) SPRAY SPRAY CONTROLLER CONTROLLER AUTO-MANUAL CONTROL AUTO-MANUA CONTROL I AUTC-MANUA_ STAION STATION (CONTROL RM( CONTROL PM) [NTROL RM: CLOSE ALL ORIFICE (SUE AT ION VALVES NOTE II (NOTE (I (NOTE C) MANUA CONTROL ETAT ON
- 1) 054)1 (L4IT Ti El)
TO TURN-ON TO VARIABLE MOE S A TE MODUL ATE CHARGING TO TURN ON ALL BACK-JR HEATER SPRAY SPRAY F LOW ALL BACK UP HEATERS CONTROL IALVE 9 VALVE HZ CON I ROL PEA (EP,S SI (SHEET i2 SIGNAL PCV -455 A PCV-4558 (SHEET 2: CIOTE 5) INOT (SHEET 12) (NOTE 6) (NCTT 6i NOT ES: I. LOG)C OUTPUT OPERATES 2 SOLENOID VENT VALVES IN SERIES TO INTERLOCK THE AIR LINE TO EACH VALVE DIAPHRAGM. THE SOLENIOD VALVES ARE DE-ENERGZED TO VENT, CAUSING THE MAIN RELIEF VALVE TO CLOSE (N 2 SECONDS.
- 2. ALL CIRCUITS CN THIS SHEET ARE NOT REDUNDANT.
- 3. LOCAL CONTROL OVERRIDES AL OTHER SIGNALS. LOCAL OVERRIDE ACTUATES ALARM IN CONTROL ROOM.
( )
- 4. PRESSURE BSTABLES PC-444B, PC-444 AND PC-445A APE ENEEGIZE NIJCLEAR STEAM SUPPLY VYCTEM TO ACTUATE.
- 5. OEN/SHUT INDICATION IN CONTRO ROOM. .S F]NCTIONA DACRAM
- 6. A LIGHT SHOULD DO PROVIDED IN THE CONTROL ROOM FOR EACH z?) cv (4! PRESSURIZER PRESSURE & LEVEL CONTROL SPRAY VALVE TO INDICATE WHEN IT IS NOT FULLY CLOSED.
UNITS I& 2 pcv 1-ks-cc o.: If Cl Cl to csos C (82 62 C) C) C Ir) IC) to tx_li Cc to VIRGINIA POWER f d 7 .-J REHISED PER OCR 2001-1268 05) u-C z*, z to NORTH ANNA POWER STATION THIS OHS SUPERSEOTS RES IOR0108t ELC C-, N A - D W -5655 D 33 C.. c_I_ REVISION DESCRIPTION COON LIII6 I 2 12-SEP-2001 061 18 LAW L L-- PCNLIi PRIOR TO USING FOR DESIGN WORK CHECK OMIS FOR WORK PENDING
QUESTIONS REPORT for NAPS 2010 RD NRC Exam rev3
- 16. O11EK1.01 016/NEW//H/4/4.1/4.4/3/
Unit 1 was initially at 100% power. A LOCA occurred, and operators have transitioned to 1-E-1, Loss of Reactor or Secondary Coolant. The following conditions exist:
- RCS pressure is 450 psig and decreasing slowly.
- SG pressures are 800 psig and decreasing slowly.
- CETCs are 580°F and decreasing.
- RCS loop cold-leg temperatures are 290°F and decreasing.
Based on these conditions, which ONE of the following identifies ALL of the systems that are providing cooling water to the core, AND includes the status of natural circulation? A. High Head SI flow AND Low Head SI flow; natural circulation is occurring B. High Head SI flow AND Low Head SI flow; natural circulation is NOT occurring C. ONLY High Head SI flow; natural circulation is occurring D. ONLY High Head SI flow; natural circulation is NOT occurring
- a. Incorrect. Plausible since this is less than the adverse pressure value that would permit stopping low head pumps (thus implying they would be injecting) however it is actually above their shutoff head.
Second part is also incorrect but plausible, temperatures decreasing tends to imply Natural Circulation however the disconnect between SG pressures and Tcs establishes that it cannot be occuring,
- b. Incorrect. First part incorrect but plausible as noted above; second part is correct break flow is the primary removal mechanism and based on the saturation temperature for the SG pressures although some reflux boiling may occur, conditions for natural circulation are not present.
- c. Incorrect. First part is correct as explained in distractor a. Second part incorrect but plausible also as discussed in Distractor a.
- d. Correct. First part is correct as explained in Distractor a; second part is also correct as discussed in distactor b.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Large Break LOCA Knowledge of the operational implications of the following concepts as they apply to the Large Break LOCA: (CFR 41.8 / 41 10 /45.3) Natural circulation and cooling, including reflux boiling Tier: 1 Group: 1 Technical
Reference:
EOP E-0, Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
NUMBER PROCEDURE TITLE REVISION 42 1 -E-O REACTOR TRIP OR SAFETY INJECTION PAGE 5 of 21 -_STEP ACTION/EXPECTEDRESPONSE I RESPONSENOTOBTAINED I 6._ VERIFY SI FLOW: C a) VERIFY HIGH-HEAD COLD LEG SI a) Verify High-Head flow indicated on the FLOW INDICATED
- following:
C
- 1-SI-Fl-1943 C
- 1-SI-FI-1943-1 OR C
- 1-Sl-Fl-1961 (NQ)
C
- 1-Sl-Fl-1962 (NQ)
C
- 1-SI-Fl-1963 (NQ)
C IF High-Head flow is NOT indicated, THEN immediately initiate ATTACHMENT 6, MANUAL VERIFICATION OF SI FLOWPATH, to restore High-Head SI flow, while continuing with this procedure. C b) Check RCS pressure LESS THAN
- C b) GO TO Step 7.
225 PSIG [450 PSIG] C c) Low-Head SI Pump flow - INDICATED c) Verify Low-Head flow indicated on the following:
-v. C
- 1-Sl-Fl-1945 C
- 1-Sl-FI-1946
.4ec C flNO1THENmanuallystartpumpsand align valves as necessary.
NUMBER ATTACHMENT TITLE ATTACHMENT 1-E-0 10 CONTINUOUS ACTION PAGE HANDOUT REVISION PAGE 42 lof5 Continuous Action Page Steps are listed on the back of this page. NOTE: The following conditions support or indicate natural circulation flow. .- NATURAL CIRCULATION VERIFICATION
- RCS subcooling based on Core Exit TCs GREATER THAN 25°F D
- SG pressures - STABLE OR DECREASING E1
- RCS Hot Leg temperatures STABLE OR DECREASING D
- Core Exit TCs STABLE OR DECREASING D
- RCS Cold Leg temperatures AT SATURATION TEMPERATURE FOR SG PRESSURE PCS Natural Circulation Display: Select from Group Display Menu RED PATH
SUMMARY
a) SUBCRITICALITY Power Range greater than 5% [Gamma-Metrics Wide-Range Power Level greater than 5 x 1001 D b) CORE COOLING Core Exit TCs greater than 1200°F OR D
- RCS Subcooling based on TCs less than 25°F [75°F] AND D
- Core Exit TCs greater than 700°F AND
- RVLIS Full Range less 48% with no RCPs running c) HEAT SINK Narrow Range level in ALL SGs less than 11% [22%] AND total Feedwater flow less than 340 gpm D d) INTEGRITY Cold leg temperature decrease greater than 100°F in last 60 minute AND ANY RCS Cold leg temperature less than Curve Limit D e) CONTAINMENT Cpntainment pressure greater than 60 psia 7 c 1 5 (3 c_ U I Vt C\
- c. 5 9 C-
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 17. 0121(2.01 01 7/BANK/NAPS/L/3/3 .3/3 .7/7/N The power to the Solid State Protection System slave relays is supplied from 1 2OVAC Vital busses.
Train A is supplied from , and Train B is supplied from A. busses I and II; busses Ill and IV B. busses I and Ill ; busses II and IV C. bus I ONLY; bus Ill ONLY D. bus II ONLY; bus IV ONLY
- a. Incorrect. Logic bays have a redundant power supply, however output bays are where the slave relays are located and powered from. The candidate who lacks detailed systems knowledge would be likely to select this distractor, since the idea of redundant power supplies would tend to make the system more robust, however it is not designed that way.
- b. Incorrect. Plausible as discussed in distractor a.
- c. Correct. As described in distractor a, the output bays have only a single power supply associated with the specific train.
- d. Incorrect. Candidate could think that the even busses are used to supply the slave relays. As noted above unless the candidate has detailed systems knowledge they can not eliminate this as a choice, as each train has two vital busses.
Reactor Protection System Knowledge of bus power supplies to the following: (CFR: 41.7) RPS channels, components, and interconnections Tier: 2 Group: 1 Technical
Reference:
1 -OP-26A, NCRODP module Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
DOMINION 1 -OP-26A North Anna Power Station Revision 50 Page 62 of 141 (Page 3 of 4) Attachment 24 1-EP-CB-04A, 120 VAC Vital Bus Distrtn Panel 1-I LOCATION: COMPUTER ROOM POWER SUPPLY: INVERTER 1-I OR SOLA
REFERENCE:
11715-FE-1AA, hA Breaker Required md No. Load Position Verifier Verifier 1-EI-CB-25: l-CH-HCV-1110, BORIC ACID TKA RECIRC VALVE CONT, AND 1-SI-HIC-1947, HYDRO TEST PP SPEED CONT (Speed Signal) AND 1-CH-HCV-1 142, RHR LETDOWN ISOL VALVE, AND 23 1-SI-HIC 1936, ACCUM N2 VENT LINE TO GW o (Controller), AND 1-LM-HIC-1O1A, CONTAIN VAC SET PT PARTIAL AIR PRESS, AND 1-SI-HIC-100, CONTAINMENT NITROGEN SUPPLY HEADER 24 Off SPARE 25 On CONTROL RM EXHAUST AIR DAMPER 2 Off SPARE 1-IA-D-2A, CONTAINMENT INSTRUMENT 27 Off AIR COMPRESSOR DRYER (Reference 2.4.4) 1-EP-CB-80E, 28 INSTRUMENTATION DISTRIBUTION PANEL 1-V o 29 On SSPS (TRAIN A OUTPUTS) 30 On SSPS AUX. RELAY RACK (TRAIN A) 31 Off SPARE 1 -EP-CB-80A, 32 On 120 VAC Instrumentation Distribution Panel 11
DOMINION 1-OP-26A North Anna Power Station Revision 50 Page 70 of 141 (Page 3 of 3) Attachment 26 1-EP-CB-04C, 120 VAC Vital Bus Distrtn Panel 1-Ill LOCATION: HATHAWAY ROOM POWER SUPPLY: INVERTER 1-Ill OR SOLA
REFERENCE:
11715-FE-1AC, 11B Breaker Required md No. Load Position Verifier Verifier 28 1-EI-CB-64B, SSPS (TRAIN B) OUTPUTS On 29 1-EI-CB-47F, SSPS TRAIN B OUTPUT On 1-IA-D-2B, CONTAINMENT INSTRUMENT 30 Off AIR DRYER (Reference 2.4.4) 1-EP-CB-80G, INSTRUMENTATION DISTRIBUTION 31 PANEL 1 -VII o 32 1-EP-CB-80C, Vital Instmment Panel 1-ITT On 33 RADIATION MONITOR CABINET 1 -2 On 34 1-EI-CB-1OC, COMPUTER CAB I/O On 35 FDR. FROM TRANSFER SW. III On
Reactor Protection Systems General System Operation also located in the Instrument Rack Room. Input relay bays I, II, III, and IV are powered from the respective 120V ac Vital Busses I, II, Ill, and IV. The Train A __-Logic Bay is powered from Vital Busses I and II. The Train B Logic Bay is powered from Vital Busses III and IV. The Train A Slave Relays are powered from Vital Bus I, and Train B Slave Relays are powered from Vital Bus III.
,/ Figure 77-10-NA and Figure 77-11-NA illustrate the power distribution to the ctor Protection Systems.
NCRODP-77-NA Page 45 10/08/07
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 18. 01 2K5.0 1 018/BANK/HARRIS 2007 NRC/L/2/3 .3/3.8/7/
Which ONE of the following reactor trips is designed to protect the core from a Departure from Nucleate Boiling condition? A. Overpower Delta T B. Overtemperature Delta T C. Pressurizer High Level D. Steam Generator Low-Low Water Level
- a. Incorrect. As discussed in the TS basis this trip serves to provide protection against violating allowable heat rate generation. Plausible since candidate may confuse this with the purpose of OTDT trip which is the correct answer.
- b. Correct. As discussed in the TS basis this trip function ensures that DNBR is maintained within design limits.
- c. Incorrect. As discussed in the TS basis this trip serves as a backup to the PRZR high pressure trip which provides RCS intergrity protection. Plausible since for some events, such as PRZR steam space leaks, PRZR water level will increase as pressure decreases, again while that event would reduce DNBR, the high level trip itself is not credited for mitigating the postulated scenario.
- d. Incorrect. As discussed in the TS basis this trip functions to protect against a loss of heat sink.
Plausible since the candidate could infer that this trip is preemtive for DNB concerns since it implies that if left unchecked RCS temperature would increase reducing DNBR, however as stated, this is NOT the TS function of the trip Reactor Protection System Knowledge of the operational implications of the following concepts as the apply to the RPS: (CFR: 41.5 /45.7) DNB Tier: 2 Group: 1 Technical
Reference:
TS bases 3.3.1, Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
NUCLEAR DESIGN IN FORMATION PORTAL RTS Instrumentation B 3.3.1 BASES APPLICABLE 5. Source Range Neutron Flux (continued) SAFETY ANALYSES, LCO, The Source Range Neutron Flux Function provides and protection for control rod withdrawal from subcritical, APPLICABILITY boron dilution and control rod ejection events. In MODE 2 when below the P-6 setpoint and in MODES 3, 4, and 5 when there is a potential for an uncontrolled RCCA bank rod withdrawal accident, the Source Range Neutron Flux trip must be OPERABLE. Two OPERABLE channels are sufficient to ensure no single random failure will disable this trip Function. Above the P-6 setpoint, the Intermediate Range Neutron Flux trip and the Power Range Neutron FluxLow Setpoint trip will provide core protection for reactivity accidents. Above the P-6 setpoint, the NIS source range detectors are de-energized and inoperable. In MODES 3, 4, and 5 with all rods fully inserted and the Rod Control System not capable of rod withdrawal, and in MODE 6, the outputs of the Function to RTS logic are not required OPERABLE. The requirements for the NIS source range detectors to monitor core neutron levels and provide indication of reactivity changes that may occur as a result of events like a boron dilution are addressed in LCO 3.9.3, Nuclear Instrumentation, for MODE 6.
- 6. Overtemperature AT The Overtemperature AT trip Function is provided to ensure that the design limit DNBR is met. This trip Function also limits the range over which the Overpower
/ / AT trip Function must provide protection. The inputs to the Overtemperature AT trip include pressurizer pressure, coolant temperature, axial power distribution, LW and reactor power as indicated by loop AT assuming full reactor coolant flow. Protection from violating the DNBR limit is assured for those transients that are slow with respect to delays from the core to the measurement system. The Function monitors both variation in power and flow since a decrease in flow has the same effect on AT as a power increase. The Overtemperature AT trip (conti nued)
North Anna Units 1 and 2 B 3.3.1-16 Revision 0
NUCLEAR DESIGN INFORMATION PORTAL RTS Instrumentation B 3.3.1 BASES APPLICABLE 6. Overtemperature AT (continued) SAFETY ANALYSES, LCO, Function uses each loops AT as a measure of reactor and power and is compared with a setpoint that is APPLICABILITY automatically varied with the following parameters:
- reactor coolant average temperaturethe trip setpoint is varied to correct for changes in coolant density and specific heat capacity with changes in coolant temperature;
- pressurizer pressurethe trip setpoint is varied to correct for changes in system pressure; and
- axial power distributionf(AI), the trip setpoint is varied to account for imbalances in the axial power distribution as detected by the NIS upper and lower power range detectors. If axial peaks are greater than the design limit, as indicated by the difference between the upper and lower NIS power range detectors, the trip setpoint is reduced in accordance with Note 1 of Table 3.3.1-1.
Dynamic compensation is included for system piping delays from the core to the temperature measurement system. The Overtemperature AT trip Function is calculated for each loop as described in Note 1 of Table 3.3.1-1. Trip occurs if Overtemperature AT is indicated in two loops. The pressure and temperature signals are used for other control functions. The actuation logic must be able to withstand an input failure to the control system, which may then require the protection function actuation, and a single failure in the other channels providing the protection function actuation. Note that this Function also provides a signal to generate a turbine runback prior to reaching the trip setpoint. A turbine runback will reduce turbine power and reactor power. Additionally, the turbine runback setpoint blocks automatic and manual rod withdrawal. A reduction in power will normally alleviate the Overtemperature AT condition and may prevent a reactor trip. The LCO requires all three channels of the Overtemperature AT trip Function to be OPERABLE. Note that the Overtemperature AT Function receives input from (conti nued) North Anna Units 1 and 2 B 3.3.1-17 Revision 0
NUCLEAR DESIGN INFORMATION PORTAL RTS Instrumentation B 3.3.1 BASES APPLICABLE 6. Overtemperature AT (continued) SAFETY ANALYSES, LCO, channels shared with other RTS Functions. Failures that and affect multiple Functions require entry into the APPLICABILITY Conditions applicable to all affected Functions. In MODE 1 or 2, the Overtemperature AT trip must be OPERABLE to prevent DNB. In MODE 3, 4, 5, or 6, this trip Function does not have to be OPERABLE because the reactor is not operating and there is insufficient heat production to be concerned about DNB.
- 7. Overpower AT The Overpower AT trip Function ensures that protection is provided to ensure the integrity of the fuel (i.e., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions. This trip Function also limits the required range of the Overtemperature AT trip Function and provides a backup to the Power Range Neutron FluxHigh Setpoint trip. The Overpower AT trip Function ensures that the allowable heat generation rate (kW/ft) of the fuel is not exceeded. It uses the AT of each loop as a measure of reactor power with a setpoint that is automatically varied with the following parameters:
- reactor coolant average temperaturethe trip setpoint is varied to correct for changes in coolant density and specific heat capacity with changes in coolant temperature; and
- rate of change of reactor coolant average temperatureincluding dynamic compensation for the delays between the core and the temperature measurement system. The function generated by the rate lag controller for Tavg dynamic compensation is represented by the expression: 3s/1-I-t t
s . The time constant utilized in the rate lag controller for Tavg is t 3 The Overpower AT trip Function is calculated for each loop as per Note 2 of Table 3.3.1-1. Trip occurs if Overpower AT is indicated in two loops. Note that this Function also provides a signal to generate a turbine runback prior to reaching the Allowable Value. A turbine runback will reduce turbine power and reactor power. (continued) North Anna Units 1 and 2 B 3.3.1-18 Revision 0
NUCLEAR DESIGN INFORMATION PORTAL RTS Instrumentation B 3.3.1 BASES APPLICABLE 7. Overpower AT (continued) SAFETY ANALYSES, LCO, Additionally, the turbine runback setpoint blocks and automatic and manual rod withdrawal. A reduction in APPLICABILITY power will normally alleviate the Overpower AT condition and may prevent a reactor trip. The LCO requires three channels of the Overpower AT trip Function to be OPERABLE. Note that the Overpower AT trip Function receives input from channels shared with other RTS Functions. Failures that affect multiple Functions require entry into the Conditions applicable to all affected Functions. In MODE 1 or 2, the Overpower AT trip Function must be OPERABLE. These are the only times that enough heat is generated in the fuel to be concerned about the heat generation rates and overheating of the fuel. In MODE 3, 4, 5, or 6, this trip Function does not have to be OPERABLE because the reactor is not operating and there is insufficient heat production to be concerned about fuel overheating and fuel damage.
- 8. Pressurizer Pressure The same sensors provide input to the Pressurizer PressureHigh and Low trips and the Overtemperature AT trip.
- a. Pressurizer PressureLow The Pressurizer PressureLow trip Function ensures that protection is provided against violating the DNBR limit due to low pressure.
The LCO requires three channels of Pressurizer PressureLow to be OPERABLE. In MODE 1, when DNB is a major concern, the Pressurizer PressureLow trip must be OPERABLE. This trip Function is automatically enabled on increasing power by the P-7 interlock (NIS power range P-1O or turbine impulse pressure greater than approximately 10% of full power equivalent (P-13)). On decreasing power, this trip Function is automatically blocked below P-7. Below the P-7 setpoint, no conceivable power distributions can occur that would cause DNB concerns. North Anna Units 1 and 2 B 3.3.1-19 Revision 0
NUCLEAR DESIGN INFORMATION PORTAL RTS Instrumentation B 3.3.1 BASES APPLICABLE 8. Pressurizer Pressure (continued) SAFETY ANALYSES, LCO, b. Pressurizer PressureHigh and APPLICABILITY The Pressurizer PressureHigh trip Function ensures that protection is provided against overpressurizing the RCS. This trip Function operates in conjunction with the pressurizer relief and safety valves to prevent RCS overpressure conditions. The LCO requires three channels of the Pressurizer PressureHigh to be OPERABLE. The Pressurizer PressureHigh LSSS is selected to be below the pressurizer safety valve actuation pressure and above the power operated relief valve (PORV) setting. This setting minimizes challenges to safety valves while avoiding unnecessary reactor trip for those pressure increases that can be controlled by the PORVs. In MODE 1 or 2, the Pressurizer PressureHigh trip must be OPERABLE to help prevent RCS overpressurization and minimize challenges to the relief and safety valves. In MODE 3, 4, 5, or 6, the Pressurizer PressureHigh trip Function does not have to be OPERABLE because transients that could cause an overpressure condition will be slow to occur. Therefore, the operator will have sufficient time to evaluate unit conditions and take corrective actions. Additionally, low temperature overpressure protection systems provide overpressure protection when below MODE 4.
- 9. Pressurizer Water LevelHigh The Pressurizer Water LevelHigh trip Function provides a backup signal for the Pressurizer PressureHigh trip and also provides protection against water relief 14 through the pressurizer safety valves. These valves are designed to pass steam in order to achieve their design U energy removal rate. A reactor trip is actuated prior to the pressurizer becoming water solid. The LCO requires three channels of Pressurizer Water LevelHigh to be OPERABLE. The pressurizer level channels are used as input to the Pressurizer Level Control System. A fourth channel is not required to address control/protection (conti nued)
North Anna Units 1 and 2 B 3.3.1-20 Revision 0
NUCLEAR DESIGN INFORMATION PORTAL RTS Instrumentation B 3.3.1 BASES APPLICABLE 9. Pressurizer Water LevelHigh (continued) SAFETY ANALYSES, LCO, interaction concerns. The level channels do not actuate and the safety valves, and the high pressure reactor trip is APPLICABILITY set below the safety valve setting. Therefore, with the slow rate of charging available, pressure overshoot due to level channel failure cannot cause the safety valve to lift before reactor high pressure trip. In MODE 1, when there is a potential for overfilling the pressurizer, the Pressurizer Water LevelHigh trip must be OPERABLE. This trip Function is automatically enabled on increasing power by the P-7 interlock. On decreasing power, this trip Function is automatically blocked below P-7. Below the P-7 setpoint, transients that could raise the pressurizer water level will be slow and the operator will have sufficient time to evaluate unit conditions and take corrective actions.
- 10. Reactor Coolant FlowLow The Reactor Coolant FlowLow trip Function ensures that protection is provided against violating the DNBR limit due to low flow in one or more RCS loops, while avoiding reactor trips due to normal variations in ioop flow.
Above the P-7 setpoint, the reactor trip on low flow in two or more RCS loops is automatically enabled. Above the P-8 setpoint, which is approximately 30% RTP, a lOSS of flow in any RCS loop will actuate a reactor trip. Each RCS loop has three flow detectors to monitor flow. The flow signals are not used for any control system input. The LCO requires three Reactor Coolant FlowLow channels per loop to be OPERABLE in MODE 1 above P-7. In MODE 1 above the P-8 setpoint, a loss of flow in one RCS loop could result in DNB conditions in the core because of the higher power level. In MODE 1 below the P-8 setpoint and above the P-7 setpoint, a loss of flow in two or more ioops is required to actuate a reactor trip because of the lower power level and the greater margin to the design limit DNBR. Below the P-7 setpoint, all reactor trips on low flow are automatically blocked since there is insufficient heat production to generate DNB conditions. North Anna Units 1 and 2 B 3.3.1-21 Revision 0
NUCLEAR DESIGN INFORMATION PORTAL RTS Instrumentation B 3.3.1 BASES APPLICABLE 11. Reactor Coolant Pump (RCP) Breaker Position SAFETY ANALYSES, LCO, Both RCP Breaker Position trip Functions operate from and three pairs of auxiliary contacts, with one pair on each APPLICABILITY RCP breaker with one contact supplying each train. These (continued) Functions anticipate the Reactor Coolant FlowLow trips to avoid RCS heatup that would occur before the low flow trip actuates. The RCP Breaker Position (Single Loop) trip Function ensures that protection is provided against violating the DNBR limit due to a loss of flow in one RCS loop. The position of each RCP breaker is monitored. If one RCP breaker is open above the P-8 setpoint, a reactor trip is initiated. This trip Function will generate a reactor trip before the Reactor Coolant FlowLow (Single Loop) trip setpoint is reached. The LCO requires one RCP Breaker Position channel per RCP to be OPERABLE. One OPERABLE channel is sufficient for this trip Function because the RCS FlowLow trip alone provides sufficient protection of unit SLs for loss of flow events. The RCP Breaker Position trip serves only to anticipate the low flow trip, minimizing the thermal transient associated with loss of a pump. This Function measures only the discrete position (open or closed) of the RCP breaker. Therefore, the Function has no adjustable trip setpoint with which to associate an LSSS. In MODE 1 above the P-8 setpoint, when a loss of flow in any RCS loop could result in DNB conditions in the core, the RCP Breaker Position (Single Loop) trip must be OPERABLE. In MODE 1 below the P-8 setpoint, a loss of flow in two or more loops is required to actuate a reactor trip because of the lower power level and the greater margin to the design limit DNBR. The RCP Breaker Position (Two Loops) trip Function ensures that protection is provided against violating the DNBR limit due to a loss of flow in two or more RCS loops. The position of each RCP breaker is monitored. Above the P-7 setpoint and below the P-8 setpoint, a loss of flow in two or more loops will initiate a reactor (conti nued) North Anna Units 1 and 2 B 3.3.1-22 Revision 0
- NUCLEAR DESIGN INFORMATION PORTAL RTS Instrumentation B 3.3.1 BASES APPLICABLE 11. Reactor Coolant Pump (RCP) Breaker Position (continued)
SAFETY ANALYSES, LCO, trip. This trip Function will generate a reactor trip and before the Reactor Coolant FlowLow (Two Loops) trip APPLICABILITY setpoint is reached. The LCO requires one RCP Breaker Position channel per RCP to be OPERABLE. One OPERABLE channel is sufficient for this Function because the RCS FlowLow trip alone provides sufficient protection of unit SLs for loss of flow events. The RCP Breaker Position trip serves only to anticipate the low flow trip, minimizing the thermal transient associated with loss of an RCP. This Function measures only the discrete position (open or closed) of the RCP breaker. Therefore, the Function has no adjustable trip setpoint with which to associate an LSSS. In MODE 1 above the P-7 setpoint and below the P-8 setpoint, the RCP Breaker Position (Two Loops) trip must be OPERABLE. Below the P-7 setpoint, all reactor trips on loss of flow are automatically blocked since no conceivable power distributions could occur that would cause a DNB concern at this low power level. Above the P-7 setpoint, the reactor trip on loss of flow in two RCS loops is automatically enabled. Above the P-8 setpoint, a loss of flow in any one loop will actuate a reactor trip because of the higher power level and the reduced margin to the design limit DNBR.
- 12. Undervoltage Reactor Coolant Pumps The Undervoltage RCPs reactor trip Function ensures that protection is provided against violating the DNBR limit due to a loss of flow in two or more RCS loops. The voltage to each RCP bus is monitored. Above the P-7 setpoint, a loss of voltage detected on two or more RCP buses will initiate a reactor trip. This trip Function will generate a reactor trip before the Reactor Coolant FlowLow (Two Loops) trip setpoint is reached. Time delays are incorporated into the Undervoltage RCPs channels to prevent reactor trips due to momentary electrical power transients.
(continued) North Anna Units 1 and 2 B 3.3.1-23 Revision 0
NUCLEAR DESIGN INFORMATION PORTAL RTS Instrumentation B 3.3.1 BASES APPLICABLE 12. Undervoltage Reactor Coolant Pumps (continued) SAFETY ANALYSES, LCO, The LCO requires three Undervoltage RCPs channels to be and OPERABLE. Each channel monitors one RCP bus voltage with APPLICABILITY two sensors. One sensor monitors from A to B phases, while the other sensor senses from the B to C phases. In MODE 1 above the P-7 setpoint, the Undervoltage RCP trip must be OPERABLE. Below the P-7 setpoint, all reactor trips on loss of flow are automatically blocked since no conceivable power distributions could occur that would cause a DNB concern at this low power level. Above the P-i setpoint, the reactor trip on loss of flow in two or more RCS loops is automatically enabled.
- 13. Underfreguency Reactor Coolant Pumps The Underfrequency RCPs reactor trip Function ensures that protection is provided against violating the DNBR limit due to a loss of flow in two or more RCS loops from a major network frequency disturbance. An underfrequency condition will slow down the pumps, thereby reducing their coastdown time following a pump trip. The proper coastdown time is required so that reactor heat can be removed immediately after reactor trip. The frequency of each RCP bus is monitored. Above the P-7 setpoint, a loss of frequency detected on two or more RCP buses will initiate a reactor trip. This trip Function will generate a reactor trip before the Reactor Coolant FlowLow (Two Loops) trip setpoint is reached. Time delays are incorporated into the Underfrequency RCPs channels to prevent reactor trips due to momentary electrical power transients.
The LCO requires three Underfrequency RCPs channels to be OPERABLE with each channel monitoring one bus. In MODE 1 above the P-7 setpoint, the Underfrequency RCPs trip must be OPERABLE. Below the P-7 setpoint, all reactor trips on loss of flow are automatically blocked since no conceivable power distributions could occur that would cause a DNB concern at this low power level. Above the P-7 setpoint, the reactor trip on loss of flow in two or more RCS loops is automatically enabled. Regarding RCP Underfrequency Testing, it should be noted that test circuits have not been installed on Unit 1, therefore, such testing can only be performed on Unit 2. North Anna Units 1 and 2 B 3.3.1-24 Revision 8
NUCLEAR DESIGN INFORMATION PORTAL RTS Instrumentation B 3.3.1 BASES APPLICABLE 14. Steam Generator Water LevelLow Low SAFETY ANALYSES, LCO, The SG Water LevelLow Low trip Function ensures that and protection is provided against a loss of heat sink and APPLICABI LITY actuates the Auxiliary Feedwater (AFW) System prior to (conti nued) p ,- uncovering the SG tubes. The SGs are the heat sink for the reactor. In order to act as a heat sink, the SGs must contain a minimum amount of water. A narrow range low low level in any SG is indicative of a loss of heat sink for the reactor. The level transmitters provide input to the SG Level Control System. Therefore, the actuation logic must be able to withstand an input failure to the control system, which may then require the protection function actuation, and a single failure in the other channels providing the protection function actuation. This Function also performs the ESFAS function of starting the AFW pumps on low low SG level. The LCO requires three channels of SG Water LevelLow Low per SG to be OPERABLE. These channels for the SGs measure level with a narrow range span. In MODE 1 or 2, when the reactor requires a heat sink, the SG Water LevelLow Low trip must be OPERABLE. The normal source of water for the SGs is the Main Feedwater (MFW) System (not safety related). The AFW System is the safety related backup source of water to ensure that the SGs remain the heat sink for the reactor. In MODE 3, 4, 5, or 6, the SG Water LevelLow Low Function does not have to be OPERABLE because the reactor is not operating or even critical. Decay heat removal is normally accomplished by Main Feedwater System or AFW System in MODE 3 and by the Residual Heat Removal (RHR) System in MODE 4, 5, or 6.
- 15. Steam Generator Water LevelLow, Coincident With Steam Flow/Feedwater Flow Mismatch SG Water LevelLow, in conjunction with the Steam Flow/Feedwater Flow Mismatch, ensures that protection is provided against a loss of heat sink. In addition to a decreasing water level in the SG, the difference between feedwater flow and steam flow is evaluated to determine if feedwater flow is significantly less than steam flow.
With less feedwater flow than steam flow, SG level will decrease at a rate dependent upon the magnitude of the difference in flow rates. There are two SG level (conti nued) North Anna Units 1 and 2 B 3.3.1-25 Revision 0
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 19. 013K2.O1 019/BANKI/L/2/3.6!3.8/2/
Which ONE of the following identifies the source of control power for the breaker for 1-SI-P-lA, A Low Head Safety Injection Pump? A. 12OVAC Vital bus 1-I B. 12OVAC Vital bus 1-Ill C. 125VDC Vital bus 1-I D. 125VDC Vital bus 1-Ill A. Incorrect. Plausible since the candidate who lacks knowledge of both the power supply to the component and the source of control power for that bus may be unsure of the source; moreover they may confuse the source of control power to the SSPS with that of the breaker control power. B. Incorrect. Plausible as discussed above. C. Correct. This is the source of control power for this piece of equipment which is powered from the H 4160v Emergency Bus. D. Incorrect. Plausible as discussed above. Engineered Safety Features Actuation System (ESFAS) Knowledge of bus power supplies to the following: (CFR: 41.7) ESFAS/safeguards equipment control Tier: 2 Group: 1 Technical
Reference:
0-AP-lO, l-OP-26A Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
NUMBER ATTACHMENT TITLE ATTACHMENT 0-AP-lO 17 LOSS OF DC BUS 1-I REVISION PAGE 62 2of4 -FJ-I ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED (I
- 6. TAKE REQUIRED ACTIONS FOR LOSS OF D BUS 1-I (1-EP-CB-12A):
D a) Verify Generator Output Breaker G-12 - C a) IF NOT open within 30 seconds of OPEN unit trip, THEN manually open 0-12. b) As required, locally operate breakers on the following busses using 0-MOP-26.1 1, 4160-VOLT BREAKER LOCAL MANUAL OPERATION, or 0-MOP-26.10, 480-VOLT BREAKER LOCAL MANUAL OPERATION, because of loss of control power: C
- 4160-Volt Bus 1C (1-EP-SW-03) c
- 4160-Volt Bus 1H (1-EE-SW-01) ryQ 0
C
- 4160-Volt Bus 2G (2-EP-SW-04)
C
- 480-Volt Bus 1C1 (1-EP-SS-07)
C
- 480-Volt Bus 1C2 (1-EP-SS-04) )Op (tLLI7 C
- 480-Volt Bus 1H (1-EE-SS-01) 0 C
- 480-Volt Bus 1H1 (1-EE-SS-03)
C
- 480-Volt Bus 1G2 (1-EP-SS-11)
C
- 480-Volt Bus 1G3 (1 -EP-SS-1 2)
C
- 480-Volt Bus 2G2 (2-EP-SS-09)
C
- Exciter Field Breaker C
- 1 5F1, 4160 Volt 1 F Transfer Bus Feed (STEP 6 CONTINUED ON NEXT PAGE)
DOMINION 1-OP-26A North Anna Power Station Revision 50 Page 29 of 141 (Page 1 of 2) Attachment 10 1-EP-CB-12A, 125V DC Bus 1-I LOCATION: EMERGENCY SWITCHGEAR ROOM POWER SUPPLY: BATT 1-1, CHGR 1-1 or ic-i
REFERENCE:
1 1715-FE-1E Breaker Required md No. Load Position Verifier Verifier Locked Off TIE TO DC BUS 1-IT CONTROL POWER TO GEN. EXCITER SWGR: 2 Generator Exciter Breaker Voltage Regulator Base On Adjuster Off SPARE 1-EP-PNL--1E57 4 n 1H EDG ROOM EMERG. LIGHTING CONTROL POWER TO 480V BUS iHi 6 On CONTROL POWER TO 480V BUS 1CT & 1C2 7 CONTROL POWER TO 4160V BUS 1C, On 1C UV Ckt, and Speed Sensing Panel for iC Bus 8 l-EP-CB-28AX, On LOAD SHED AUX RELAY CAB 1-EP-CB-200, 9 G-12BRKRCONTPNL on 1-EP-CB-25A, 10 DCPNL1-IV on 11 Locked Off FEEDER FROM SWING CHGR 12 FEEDER NORMAL CHARGER 2-EP-CB-002, 13 APP R CAB AND INVERTER 1-I CONTROL POWER TO 4160V BUS IH, 14 1H UV Ckt, and EDG Auto Start
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 20. 01 3K4.06 020/MODIFIED/NAPS ID:50 1 78/H/3/4.0/4.3/2/
Given the following conditions:
- Unit 1 was initially at 100% power.
- Reactor trip and SI occurred due to a large-break LOCA.
- A loss of inventory outside containment is occurring and the crew is unable to isolate it.
- RWST level is continuing to decrease.
Which ONE of the following identifies the RWST level setpoint at which automatic swapover to Containment Sump will occur AND includes the actions that will prevent the automatic swapover? A. 16%; reset CDA and push both SI Recirc Mode Reset Pushbuttons B. 16%; reset SI and push both SI Recirc Mode Reset Pushbuttons C. 23%; reset CDA and push both SI Recirc Mode Reset Pushbuttons D. 23%; reset SI and push both SI Recirc Mode Reset Pushbuttons
- a. Incorrect. First part is correct. Second part is incorrect but plausible since transfer to cold leg recirc is usually associated with DBA event like a LBLOCA where CDA will actuate. The candidate who lacks detailed knowledge of the ESFAS may conclude that this is logical and default to this distractor.
Further, CDA reset is an action taken to shutdown/prevent auto-start of certain ESF equipment such as Inside Recirc Spray Pumps, thus the candidate could also confuse this relationship.
- b. Correct. First part is correct as noted above; second part is also correct--the SI signal establishes the permissive, which is maintained until the associated train SI Recirc Mode Reset Pushbutton is depressed; the SI signal must be reset before pushing the SI Recirc Mode Reset Pushbutton will have an effect.
- c. Incorrect. First part is incorrect but plausible since this is the setpoint at which the EOP directs transition to ES-i .3, Transfer to Cold Leg Recirculation; second part incorrect but plausible as discussed in Distractor a.
- d. Incorrect. First part is incorrect as discussed in Distractor c; second part is correct as explained in answer.
QUESTIONS REPORT for NAPS 2010 RD NRC Exam rev3 Engineered Safety Features Actuation System (ESFAS) Knowledge of ESFAS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) Recirculation actuation system reset Tier: 2 Group: 1 Technical
Reference:
TS 3.3.2 tbl 3.3.2-1, TR 4.6 tbl 4.6-11, EOP E-1, AR J-A2, 1-ECA-1.1 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
NUCLEAR DESIGN INFORMATION PORTAL ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 4 of 4) Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE
- 6. Auxiliary Feedwater
- a. Automatic Actuation Logic 1, 2, 3 2 trains 6 SR 3.3.2.2 NA and Actuation Relays SR 3.3.2.3 SR 3.3.2.5
- b. SG Water LevelLow Low 1, 2, 3 3 per SG 0 SR 3.3.2.1 17%
SR 3.3.2.4 SR 3.3.2.8 SR 3.3.2.9
- c. Safety Injection Refer to Function 1 (Safety Injection) for all initiation functions and requirements.
- d. Loss of Offsite Power 1, 2, 3 1 per bus, F SR 3.3.2.6 2184 V 2 buses SR 3.3.2.8 SR 3.3.2.9
- e. Trip of all Main Feedwater 1, 2 2 per pump H SR 3.3.2.7 NA Pumps SR 3.3.2.9
- 7. Automatic Switchover to Containment Sump
- a. Automatic Actuation Logic 1, 2, 3, 4 2 trains C SR 3.3.2.2 NA and Actuation Relays SR 3.3.2.3 SR 3.3.2.5
- b. RWST LevelLow Low 1, 2, 3, 4 4 I SR 3.3.2.1 15% and SR 3.3.2.4 17%
SR 3.3.2.8 SR 3.3.2.9 Coincident with Safety Refer to Function 1 (Safety Injection) for all initiation functions and Injection requirements.
- 8. ESFAS Interlocks
- a. Reactor Trip, P-4 1, 2, 3 1 per train, F SR 3.3.2.7 NA 2 trains
- b. Pressurizer Pressure, Pli 1, 2, 3 3 J SR 3.3.2.1 2010 psig SR 3.3.2.8 Low Low, Pi2
- c. T 9 1, 2, 3 1 per loop J SR 3.3.2.1 542°F and SR 3.3.2.8 545°F North Anna Units 1 and 2 3.3.2-11 Amendments 250/230
NUCLEAR DESIGN INFORMATION PORTAL ESFAS Instrumentation Trip Setpoints 4.6 Table 4.6-1 (page 3 of 3) ESFAS Instrumentation Trip Setpoints FUNCTION TRIP SETPOINT
- 6. Auxiliary Feedwater Pump Start
- a. Manual Initiation N.A.
- b. Automatic Actuation Logic N.A.
- c. Steam Generator Water 18% of narrow range instrument span each LevelLow-Low steam generator
- d. Safety Injection All Safety Injection Setpoints
- e. Loss of Offsite Power 2392 volts on Transfer Bus
- f. Trip of Main Feed Pump N.A.
- 7. Automatic Switchover to Containment Sump
- a. Automatic Actuation Logic N.A.
and Actuation Relays
- b. RWST LevelLow Low 16.0% +/- 1.0%
NAPS TRM 4.6-4 Rev 64, 03/27/07
CONTINUOUS ACTION PAGE FOR 1-E-1
- 1. ADVERSE CONTAINMENT CRITERIA IF either of the following conditions exist, THEN use setpoints in brackets:
C
- 20 psia Containment pressure, OR C
- Containment radiation has reached or exceeded 1 .0E5 R/hr (70% on High Range Recorder).
- 2. RCP TRIP CRITERIA IF both conditions listed below exist, THEN trip all RCPs:
C Charging Pumps AT LEAST ONE RUNNING AND FLOWING TO RCS, AND C
- RCS subcooling based on Core Exit TCs LESS THAN 25°F [85°F].
- 3. CHARGING PUMP RECIRC PATH CRITERIA C
- IF RCS pressure decreases to less than 1275 psig [1475 psig] AND RCPs tripped, THEN close Charging Pump Recirc Valves.
C
- IF RCS pressure increases to 2000 psig, THEN open Charging Pump Recirc Valves.
- 4. SI REINITIATION CRITERIA IF either condition listed below occurs, THEN manually start Charging Pumps and align BIT:
C
- RCS subcooling based on Core Exit TCs LESS THAN 25°F [75°F], OR C
- PRZR level CANNOT BE MAINTAINED GREATER THAN 21% [26%].
- 5. ECST LEVEL CRITERIA C WHEN the ECST level decreases to 40%, THEN initiate 1-AP-22.5, LOSS OF EMERGENCY CONDENSATE STORAGE TANK 1 -CN-TK-1.
- 6. SECONDARY INTEGRITY CRITERIA IF either of the following conditions exist AND the affected SG has NOT been isolated, THEN GO TO 1-E-2, FAULTED STEAM GENERATOR ISOLATION, STEP 1:
C
- Any SG pressure is decreasing in an uncontrolled manner, OR C
- Any SG has completely depressurized.
- 7. 1-E-3 TRANSITION CRITERIA IF either of the following conditions exist, THEN manually start Charging Pumps, align BIT, and GO TO i-E-3, STEAM GENERATOR TUBE RUPTURE, STEP 1:
C
- Any SG level is increasing in an uncontrolled manner, OR C
- Any SG has abnormal radiation.
- 8. COLD LEG RECIRCULATION TRANSFER CRITERIA f-r/2- -r C H RWST level decreases to less than 23%, THEN GO TO 1-ES-i .3, TRANSFER TO COLD LEG RECIRCULATION, STEP 1.
- 9. QS TERMINATION CRITERIA C WHEN RWST level is less than 3% AND QS Pump amps are FLUCTUATING, THEN perform ATTACHMENT 3, TERMINATION OF QUENCH SPRAY.
- 10. CASING COOLING TANK LEVEL C WHEN the Casing Cooling Tank level decreases to 4%, THEN ensure CDA is reset, close i-RS-MOV-100A and i-RS-MOV-iOOB and stop both Casing Cooling Pumps.
ii. RCP CRITERIA C Seal injection flow should be maintained to all RCPs.
- 12. REACTIVITY CONTROL CRITERIA C An Operator should be sent to locally close and lock 1-CH-217, PG to Blender Isolation Valve.
VIRGINIA POWER l-EI-CB-21J ANNUNCIATOR A2 l-AR-J-A2 NORTH ANNA POWER STATION REV. 0 APPROVAL: ON FILE Effective Date:06/13/97 RWS T LO RWST < 22.8% LEVEL 1.0 Probable Cause 1.1 Failure of lQSLT100A or bOB or lQSLAL100A3 or 100B3 1.2 CDA in progress 1.3 Filling Reactor Cavity during refueling 1.4 Rupture of RP system piping while aligned to RWST 2.0 Operator Action 2.1 Verify RWST level. 2.2 IF CDA in progress, THEN GO TO l-ESl.3, Transfer To Cold Leg Recirc. 2.3 IF filling Reactor Cavity using LHSI Pump, THEN ensure SI Recirc Mode Reset pushbuttons, Train A and Train B, have been depressed, to prevent swapover of LHSI Pump suction to the Containment Sump. 2.4 IF borating or filling the RWST AND level begins to decrease, THEN secure fill lineup through RP system, UNTIL RP system is verified intact. 2.5 If level is NOT low, THEN submit Work Request on level transmitter. 3.0 References 3.1 1l715LSK295B 3.2 NAPS Instrumentation Manual page QSOO3 and Q5004 3.3 Loop Diagram l-L-QS100A 3.4 EWR 89571 3.5 DCP 9013, Steam Generator Replacement 3.6 ET CE96014, Rev 0, Mode 5 & 6 Compensatory Measures Recommended for Problem Reported in Deviation Report No. N96-0278 4.0 Actuations 4.1 lQSLSLbOOA2 and 1QSLAS--100B2 RWST low level alarm switches. Feed signals to 1-QS-LAL-100A-3 and l-QSLAL-100B-3
NUMBER PROCEDURE TITLE REVISION 16 1 -ECA-1 1 LOSS OF EMERGENCY COOLANT RECIRCULATION
. PAGE 2 of 34 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
-j_STEP CAUTION:
- If emergency coolant recirculation capability is restored to at least one train during this procedure, then further recovery actions should continue by returning to the procedure and step in effect.
- If the suction source is lost to any SI Pump or Recirc Spray Pump, then the pump should be stopped.
NOTE:
- IF Containment Sump Blockage has occurred, THEN FRs should NOT be implemented until directed in this procedure.
- ATTACHMENT 2, MINIMUM SI FLOW RATE VERSUS TIME AFTER TRIP, provides adequate injection flow required.
- Setpoints in brackets [] are for adverse Containment atmosphere (20 psia Containment pressure or Containment Radiation has reached or exceeded 1 .0E5 R/hr or 70% on High Range Recorder).
- 1. CHECK EMERGENCY COOLANT Try to restore at least one train of RECIRCULATION EQUIPMENT - Emergency Coolant Recirculation AVAILABLE: Equipment:
D
- Low-Head SI Pumps EJ
- Local operations D
- Low-Head SI Pump Suction Valves from
- Electrical restoration Containment EJ
- Equipment repair RESET BOTH TRAINS OF SI IF D Perform 1-AP-0, RESETTING SI LOCALLY, NECESSARY while continuing with this procedure.
PUSH BOTH SI RECIRC MODE RESET BUTTONS CHECK RWST LEVEL GREATER THAN
- D Implement FRs as applicable.
8%
/ D GOTOStep3O.
/ c
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 21. 0 14A2.02 02 1/MODIFIED/NAPS ID:6006/H/4/3 .1/3.6/1 /
Which ONE of the following identifies the normal power source to Unit I Individual Rod Position Indication (IRPI) System, [Q includes the system response in the event of a loss of that power supply? A. MCC1H1-1; The Unit 1 IRPI System will have power for an indefinite period of time. B. MCC1H1-1; The Unit 1 IRPI System will have power for a limited period of time. C. MCC1J1-1; The Unit 1 IRPI System will have power for an indefinite period of time. D. MCC1J1-1; The Unit 1 IRPI System will have power for a limited period of time.
- a. Incorrect. First part is correct. Second part incorrect but plausible since this system unlike most relies solely on a backup battery which will provide indication for some period of time, and unlike a CVT it is not unlimited, but is a function of battery capacity, which is limited.
- b. Correct. First part is correct as mentioned above. Second part is also correct; as explained in distractor a the only backup is from a battery source and therefore has a finite life.
- c. Incorrect. First part is incorrect but plausible because the candidate who lacks detailed systems knowledge may know that vital power supplies the regulating transformer, but may not know which train. Second part is incorrect but plausible as discussed in distractor a.
- d. Incorrect. First part is incorrect but plausible as discussed in distractor c. Second part is correct as discussed above.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Rod Position Indication System (RPIS) Ability to (a) predict the impacts of the following malfunctions or operations on the RPIS; and (b) based on those on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5/43.5 /45.3/45.13) Loss of power to the RPIS Tier: 2 Group: 2 Technical
Reference:
AR A-G6, 1-OP-26A Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info: considered a KA match since the candidate must demonstrate knowledge of both the normal power supply and the limitations associated with the backup source; since NAPS does not use an IRPI inverter with CVT backup like some stations, there really is no other strategy for a loss of normal power, other than re-energizing the affected bus which makes it difficult to write a distractor that would seem plausible.
VIRGINIA POWER 1-EI-CB-21A ANNUNCIATOR G6 1-AR-A-G6 NORTH ANNA POWER STATION REV. 2 APPROVAL: ON FILE Effective Date:03/19/03 RPI SYSTEM AUX DC POWER ON NOTE: Power supplies automatically backup each other and loss of one positive and/or one negative supply does not affect system. NOTE: A battery backup is connected to the voltage regulating converter. This should provide power to the IRPI5 for a minimum of 30 minutes in the event that power is lost to the 11-111 480 volt MCC. 1.0 Probable Cause 1.1 Failure of 1-EE-VREG-2, RPI Transformer 86 Voltage Regulator 1.2 Failure of any of the four D.C. power systems 2.0 Operator Action 2.1 Check if IRPIs are indicating, IF YES, THEN GO TO Step 2.3. 2.2 IF IRPI5 are not indicating, THEN do the following:
- a. Place rods in manual.
- b. Declare IRPIs inoperable and take actions of TS 3.1.7.
- c. Check status of l-EE-VREG2, RPI Transformer 86 Voltage Regulator, as follows:
- IF NO lights lit, THEN check power supply to voltage regulator.
- LINE and INV lights should be Lit all others out.
- IF any other Light Lit, THEN secure regulator by pushing the ON-OFF push button securing the regulator.
- d. Allow unit to power down then restart unit by pushing the ON-OFF push button.
- e. Verify the unit starts and the LINE and INV lights are LIT.
IF any other light LIT, THEN submit WR and contact I&C Shop.
- f. Verify IRPIs indication is restored in Control Room.
- g. Verify that annunciator has cleared. IF not, THEN GO TO Step 2.3.
- h. Clear action of TS 3.1.7 and exit procedure.
2.3 Check power failure lamps on RPI rack to determine which power supply failed. 2.4 Submit WR on failed power supply and notify I&C Shop. 2.5 WHEN repaired by I&C, THEN clear action.
3.0 References 3.1 Westinghouse manual rod position indication system 3.2 P1 N20020030, Unit 1 Entered TS 3.0.3, Loss of all IRPI Indications 3.3 DCP 01113, RPI Voltage Regulating Transformer Replacement, Unit 1 4.0 Actuation 4.1 Internal to power supply system
DOMINION 1-OP-26A North Anna Power Station Revision 50 Page 19 of 141 (Page 3 of 3) Attachment 5 1-EE-MCC-IHI-1, IHI-1 Motor Control Center 1-EP-MC-1O LOCATION: EMERGENCY SWITCHGEAR ROOM POWER SUPPLY: 1-EE-BKR 14H-4
REFERENCE:
1 1715-FE-1Z Breaker Required md No. Load Position Verifier Verifier 1-Sw-P-b, I I In Service On G4 Rad Mon Sample Pump for I 1-SW-I-1O8 Ckt Bkr (Ref. 2.3.26) Out of Service Off 1 -EP-IV-02, H1L On Computer Transfer Inverter Circuit Breaker (SOLA) 1 -EP-CB- 1 6A, H1R On 1A Semi-Vital Bus Distribution Pnl Ckt Bkr 1-EI-CB-41B, H2L On Rod Position Indicator Rack No 2 Cabinet Ckt Bkr 1 -HV-FL-9 H2R On Filter Heater Coil Circuit Breaker l-SW-MOV-1 19, H3 On Scm Wash Pps Mkup to SW Sply Hdr No 1 Isol Vv CB H4L SPARE, H4L Off Spare Circuit Breaker H4R On Vital Bus 1 -II Bypass Transformer Circuit Bkr
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 22. O1SAK2.08 022/BANKI/H/3/2.6/2.6/4/
Unit 1 is at 100% power. A total loss of CC flow has occurred, and the crew entered 1-AP-15, Loss of Component Cooling. Which ONE of the following identifies the component that will reach its temperature limit the soonest for this event, JQ includes the maximum temperature that is allowed by 1 -AP-1 5 before a reactor trip is required? A. RCP pump radial bearing; 195°F B. RCP pump radial bearing ; 225°F C. RCP motor bearings; 195°F D. RCP motor bearings ; 225° F
- a. Incorrect. Plausible since this parameter is specifically monitored in 1-AP-15, however based on the plant conditions pump bearing temperatures should not be elevated since seal injection would still be in-service; also temperature is for motor bearing trip criteria not pump bearing, but again the two could be confused.
- b. Incorrect. First part incorrect but plausible as noted above; second part is correct.
- c. Correct. Motor bearings will heatup and reach trip criteria before other RCP temperatures. Second part is the correct trip criteria lAW 1-AP-15.
- d. Incorrect. First part is correct as discussed in answer c; second part is incorrect, but as mentioned is plausible since it is trip criteria for the pump bearing and the two could be confused.
Reactor Coolant Pump (RCP) Malfunctions Knowledge of the interrelations between the Reactor Coolant Pump Malfunctions (Loss of RC Flow) and the following: (CFR 41.7 / 45.7) CCWS Tier: 1 Group: 1 Technical
Reference:
1-AP-15 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
Doii NORTH ANNA POWER STATION ABNORMAL PROCEDURE NUMBER PROCEDURE TITLE REVISION 21 I -AP-I 5 LOSS OF COMPONENT COOLING (WITH SIX ATTACHMENTS) PAGE 1 of 9 PURPOSE To provide the actions to take in the event of a loss of Unit 1 Component Cooling System. ENTRY CONDITIONS This procedure is entered when any of the following conditions exist:
- Annunciator Panel G A-i, CC SURGE TK HI-LO LEVEL, is LIT, or
- Annunciator Panel G F-5, COMP COOL PP IA AUTO TRIP, is LIT, or
- Annunciator Panel G E-8, COMP COOL PP 18 AUTO TRIP, is LIT, or
- Annunciator Panel G B-3, CC HX iA-lB CC OUTLET LO FLOW, is LIT, or
- Annunciator Panel G C-3, CC HX OUTLET LO PRESS, is LIT, or
- Excess Letdown Heat Exchangers low flow/high temperature, or
- Non-regenerative Heat Exchangers high temperature, or
- Reactor Coolant Pumps low flow/high temperature, or
- Loss of Service Water System.
CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 21 I -AP-I 5 LOSS OF COMPONENT COOLING PAGE 5 of 9 ACTION! EXPECTED RESPONSE I I RESPONSE NOT OBTAINED
- 10 MONITOR RCP Do the following while continuing with this TEMPERATURES: procedure:
- Motor bearing temperature -
cvD a) GO TO 1-E-0, REACTOR TRIP OR SAFETY LESS THAN 195°F INJECTION. Pump radial bearing D b) WHEN Reactor is tripped, THEN stop affected temperature LESS THAN
- RCPs.
225°F Stator winding temperature - LESS THAN 300 °F
- 11. ISOLATE LETDOWN BY CLOSING D Close 1-CH-LCV-1460A by placing control switch THE FOLLOWING VALVES: in ISO.
a) Letdown Orifice Isolation Valves: l
- 1-CH-HCV-1200A D
- 1-CH-HCV-1200B D
- 1-CH-HCV-1200C b) Letdown Isolation Valves:
D
- 1-CH-LCV-1460A D
- 1-CH-LCV-1460B 12._ CHECK EXCESS LETDOWN - D Secure Excess Letdown using 1-OP-8.5, SECU RED OPERATION OF EXCESS LETDOWN.
13._ CLOSE 1-CH-FCV-1 122, CHARGING D Close 1-CH-MOV-1289A, Normal Charging Line FLOW CONTROL VALVE Isolation Valve. 14._ CLOSE 1-CH-MOV-1380, SEAL D Close 1-CH-MOV-1381, Seal Water Return WATER RETURN ISOLATION VALVE Isolation Valve.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 23. 01 5K2.0 1 023/MODIFIED/NAPS ID:71 61 /H/4/3 .3/3.7/7/
Given the following conditions:
- Unit 1 is in mode 3.
- Power Range channel N-44 is tagged out.
- A loss of 12OVAC Vital bus I-Ill occurs.
Based on these conditions, which ONE of the following identifies the response of the Source-range Nls? Source-range channel(s) A. N-31 jy will be de-energized due to the P-ic permissive. B. N-31 and N-32 will be de-energized due to the P-b permissive. C. N-31 Qpjy will be de-energized due to the P-6 permissive. D. N-31 and N-32 will both be de-energized due to the P-6 permissive.
- a. Incorrect. Plausible since the candidate who lacks detailed knowledge of train power supplies (separate P-ic logic is contained in each train of SSPS) may consider that only N-31 would be energized, erroneously attributing the change of state in P-iO to the loss of power that would be associated with B train. N-44 is tagged out so its bistables are tripped, thus when I-Ill is lost a trip is initiated on N-43 and now 2/4 PR channels are> 10% so the P-ic permissive will perform its ancillary function of de-energizing the source range detector high voltage.
- b. Correct. As discussed previously since power above P-i 0 is satisfied by virtue of the power range bistable logic, both source range detectors will be deenergized.
- c. Incorrect. Plausible because the candidate who lacks detailed knowledge of power supplies and permissives may consider that a loss of 1-Ill would cause a change in state of the P-6 bistable and attribute this as the cause.
- d. Incorrect. Plausible as discussed in Distractor c.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Nuclear Instrumentation System Knowledge of bus power supplies to the following: (CFR: 41.7) NIS channels, components, and interconnections Tier: 2 Group: 2 Technical
Reference:
UFSAR Sec. 7.2, AR5 P-D2, P-D1, A-Al, dwgs 5655D33 sh. 3 of 16 & 4 of 16 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info: K/A intent is met since question tests knowledge of power supplies and cause/effect relationship with SSPS.
NUCLEAR DESIGN INFORMATION PORTAL Revision 45.05Updated Online 03/15/10 NAPS UFSAR 7.2-4 7.2.1.1.1 Nuclear Overpower Trips 7$? \ The specific trip functions generated are as follows:
- 1. Power range high-neutron-flux tripThe power range high-neutron-flux trip circuit trips the reactor when two of the four power range channels exceed the trip setpoint.
There are two independent hi-stables, each with its own trip setting used for a high and a low setting. The high trip setting provides protection during normal power operation and is always active. The low trip setting, which provides protection during startup, can be manually bypassed when two out of the four power range channels read above approximately 10% power (P-b). Three out of the four channels below 10% automatically reinstate the trip function. Refer to Table 7,2-3 for a listing of all reactor trip system interlocks.
- 2. Intermediate range high-neutron-flux tripThe intermediate range high-neutron-flux trip circuit trips the reactor when one out of the two intermediate range channels exceeds the trip setpoint. This trip, which provides protection during reactor startup, can be manually blocked if two out of the four power range channels are above approximately 10% power (P-b).
Three out of the four power range channels below this value automatically reinstate the intermediate range high-neutron-flux trip. The intermediate range channels (including detectors) are separate from the power range channels. The intermediate range channels can be individually bypassed at the nuclear instrumentation racks to permit channel testing during plant shutdown or before startup. This bypass action is annunciated on the control board.
- 3. Source range high-neutron-flux tripThe source range high-neutron-flux trip circuit trips the reactor when one of the two source range channels exceeds the trip setpoint. This trip, which provides protection during reactor startup and plant shutdown, can be manually bypassed when one of the two intermediate range channels reads above the P-6 setpoint value and is automatically reinstated when both intermediate range channels decrease below the P-6 value. This trip is also automatically bypassed by two-out-of-four logic from the power range interlock (P-b). This trip function can also be reinstated below P-b by an administrative action requiring manual actuation of two control board mounted switches.
Each switch will reinstate the trip function in one of the two protection logic trains. The source range trip point is set between the P-6 setpoint (source range cutoff flux level) and the maximum source range flux level. The channels can be individually bypassed at the nuclear instrumentation racks to permit channel testing during plant shutdown or before startup. This bypassing action is annunciated on the control board.
- 4. Power range neutron flux rate trips (PRRT)Refer to Figure 7.2-3. The functional diagram shown includes reactor trip logic provided to trip the reactor when an abnormal rate of increase or decrease in nuclear power occurs in two out of four power range channels.
NUCLEAR DESIGN INFORMATION PORTAL Revision 45.05Updated Online 03/15/10 NAPS UFSAR 7.2-38 Table 7.2-3 REACTOR TRIP SYSTEM INTERLOCKS Designation Derivation Function Power Escalation Permissives P-6 1/2 neutron flux (intermediate Allows manual block of source range range) above setpoint reactor trip 2/2 neutron flux (intermediate Defeats the block of source range range) below setpoint reactor trip P-i 0 2/4 neutron flux (power range) Allows manual block of power range above setpoint (low setpoint) reactor trip Allows manual block of intermediate range reactor trip and intermediate range rod stops (C-i) Blocks source range reactor trip (back-up for P-6) 3/4 neutron flux (power range) Defeats the block of power range (low below setpoint setpoint) reactor trip Defeats the block of intermediate range reactor trip and intermediate range rod stops (C-i) Input to P-7 Blocks of Reactor Trips P-7 3/4 neutron flux (power range) Blocks reactor trip on low flow or below setpoint (from P-b) and 2/2 reactor coolant pump breakers open in turbine impulse chamber pressure more than one loop, undervoltage, below setpoint (from P-13) underfrequency, pressurizer low pressure, and pressurizer high level P-8 3/4 neutron flux (power range) Blocks reactor trip on low flow or below setpoint reactor coolant pump breaker open in a single loop and on turbine trip P-i 3 2/2 turbine impulse chamber Input to P-7 pressure below setpoint
NUCLEAR DESIGN INFORMATION PORTAL Revision 45.05Updated Online 03/15/10 NAPS UFSAR 7.2-11 7.2.1.1.7 Safety Injection Signal Actuation Trip A reactor trip occurs when the safety injection system is actuated. The means of actuating the safety injection system are described in Section 7.3. This trip protects the core during a loss of reactor coolant or steam-line break. Figure 7.2-9 shows the logic for this trip. A detailed functional description of the process equipment associated with this trip function is provided in Reference 3. 7.2.1.1.8 Manual Trip The manual trip consists of two redundant switches with multiple outputs on each switch. One output is used to actuate the train A trip breaker and another output actuates the train B trip breaker. Operating a manual trip switch removes the voltage from the undervoltage trip coil and energizes the shunt trip coil, either of which will cause a reactor trip. There are no interlocks that can block this trip. Figure 7.2-3 shows the manual trip logic. 7.2.1.2 Reactor Trip System Accuracies and Response Times The system accuracies and the system response times of the instrument trip signals required for plant safety are given in Tables 7,2-2 and 15.1-3, respectively. Periodic response time testing of the reactor trip and ESF systems has been established in the Technical Specifications to meet the intent of IEEE Std 338-1971. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC. The measured or verified channel response times are compared with those used in the safety evaluations. In accordance with Technical Specifications, the response times are required to be less than or equal to the times used in the safety analyses. 7.2.1.3 Reactor Trip System Interlocks 7.2.1.3.1 Power Escalation Permissives The overpower protection provided by the out-of-core nuclear instrumentation consists of three discrete, but overlapping, levels. The continuation of startup operation or power increase requires a permissive signal from the high-range instrumentation channels before the lower range level trips can be manually blocked by the operator. A one-of-two intermediate range permissive signal (P-6) is required before source range level trip blocking and detector high-voltage cutoff. Source range level trips are automatically reactivated and high voltage restored when both intermediate range channels are below the permissive (P-6) level. There is a manual reset switch for administratively reactivating the source
NUCLEAR DESIGN INFORMATION PORTAL Revision 45.05Updated Online 03/15/10 NAPS UFSAR 7.2-12 range level trip and detector high voltage when between the permissive P-6 and P- 10 level if required. Source range level trip block and high-voltage cutoff are always maintained when above the permissive P-10 level. The intermediate range level trip and power range (low setpoint) trip can only be blocked after satisfactory operation and permissive information are obtained from two out of four power range channels. Individual blocking switches are provided so that the low-range power range trip and intermediate range trip can be independently blocked. These trips are automatically reactivated when any three of the four power range channels are below the permissive (P-b) level, thus ensuring automatic activation to more restrictive trip protection. The development of permissives P-6 and P-10 is shown in Figure 7.2-10. All of the permissives are digital; they are derived from analog signals in the nuclear power range and intermediate range channels. See Table 7.2-3 for the list of reactor trip system interlocks. 7.2.1.3.2 Blocks of Reactor Trips at Low Power Interlock P-7 blocks a reactor trip at low power (below approximately 10% of full power) on a low reactor coolant flow or reactor coolant pump open breaker signal in more than one loop, reactor coolant pump undervoltage, reactor coolant pump underfrequency, pressurizer low pressure, or pressurizer high water level. See Figures 7.2-5, 7.2-6 and 7.2-8 for permissive applications. The low-power signal is derived from three out of four power range neutron flux signals below the setpoint in coincidence with two out of two turbine impulse chamber pressure signals below the setpoint (low plant load). The P-8 interlock blocks a reactor trip when the plant is below approximately 30% of full power, on a low reactor coolant flow in any one ioop, a reactor coolant pump breaker open signal in any one ioop, or turbine trip signal. Below the P-8 setpoint, the reactor will not trip with a turbine trip, or with one inactive ioop. The reactor could be allowed to operate with one inactive loop, provided Technical Specifications are amended to authorize this mode of operation. See Figure 7.2-10 for the derivation of P-8 and Figures 7.2-5 and 7.2-8 for applicable logics. See Table 7.2-3 for the list of protection system blocks. 7.2.1.4 Coolant Temperature Sensor Arrangement Three thermowell mounted resistance temperature detectors are installed in the hot leg of each ioop near the inlet to the steam generator for reactor protection and control. One thermowell mounted resistance temperature detector is installed in the cold leg of each ioop at the discharge of the reactor coolant pump for reactor protection and control.
VIRGINIA POWER 1-EI-CB--21P ANNUNCIATOR D2 1-AR-P--D2 NORTH ANNA POWER STATION REV. 0 APPROVAL: ON FILE Effective Date:11/26/97 Pb PERM 2/4 Power Range PR > 10% neutron flux BLK NIS LP TRIPS detectors sensing power equivalent to at least 10 % of full power during a Reactor startup 1.0 PROBABLE CAUSE This annunciator indicates that the Reactor has reached a power level of at least 10 percent (as indicated by 2/4 Power Range neutron flux detectors) during a Reactor startup.
2.0 REFERENCES
2.1 NA-DW-5655D33, Sheet 3, Functional Diagrams, Nuclear Instrument and Manual Trip Signals, Units 1 and 2 2.2 NA-DW-5655D33, Sheet 4, Functional Diagrams, Nuclear Instr. Permissives and Blocks, Units 1 and 2 2.3 11715ESIc1OCAF, TripPermissive and Bypass Status Lights, Sheet 6 of 8 2.4 Westinghouse PWR Systems Division, Nuclear Instrumentation System, Technical Manual 3.0 ACTUATION Bistable card NC308, of 2/4 Power Range neutron flux detectors N-41, N-42, N43, and N44, sensing Reactor power equivalent to at least 10 percent
VIRGINIA POWER 1-EI-CB-21P ANNUNCIATOR Dl 1-AR-P-D1 NORTH ANNA POWER STATION REV. 0 APPROVAL: ON FILE Effective Date:11/26/97 P-6 PERM 1/2 Intermediate IR > 1010 Range neutron flux BLK SR TRIP detectors sensing power equivalent to at least 1010 amps 1.0 PROBABLE CAUSE r) fv&ch Reactor has reached a power level equivalent to 1010 amps (as indicated by 1/2 Intermediate Range neutron flux detectors) during a Reactor startup
2.0 REFERENCES
2.1 NA-DW5655D33, Sheet 3, Functional Diagrams, Nuclear Instrument and Manual Trip Signals, Units 1 and 2 2.2 NA-DW-5655D33, Sheet 4, Functional Diagrams, Nuclear Instr. Permissives and Blocks, Units 1 and 2 2.3 11715ESK1OCAF, Trip-Permissive and Bypass Status Lights, Sheet 6 of 8 2.4 Westinghouse PWR Systems Division, Nuclear Instrumentation System, Technical Manual 3.0 ACTUATION Bistable card NC205, of Intermediate Range neutron flux detector N-35 or N36, sensing Reactor power equivalent to 1010 amps
VIRGINIA POWER l-EI-CB-21A ANNUNCIATOR Al 1-AR-A-Al NORTH ANNA POWER STATION REV. 1 APPROVAL: ON FILE Effective Date:lO/16/O1 NIS SR LOSS OF DET VOLTAGE NOTE: Annunciator is automatically disabled above Pb . / -r L N 1.0 Probable Cause 1.1 Loss of high voltage to either source range detector 1.2 Manual blocking of source range at P6 1.3 Reduction in power level below PlO with source range detectors NOT yet energized 2.0 Operator Action 2.1 Actuation due to manual blocking of source range requires no operator action. 2.2 Actuation due to Rx power decreasing below PlO requires no operator action. 2.3 Observe Source Range instruments for alarm light to determine which instrument gave alarm. 2.4 IF instrument has failed, THEN GO TO l-AP-4.l, Malfunction of Source Range Nuclear Instrumentation. 3.0 References 3.1 lAP--4.1, Malfunction Of Source Range Nuclear Instrumentation 3.2 11715ESK1OAAJ 3.3 W drawing 605lDl9 sheet 4 and 27lC35l sheet 27 3.4 Tech Spec 3.3.1.1, 3.9.2 (ITS 3.3.1 and 3.9.3) 3.5 UFSAR Chapter 7, Figure 7.23 4.0 Actuation 4.1 Source range detector N3l, or N32, loss of detector voltage comparator NC1O4 in NIS drawer N31 and N32.
FROM R BLOCS. RCH I/N 3EA LOGIC (SHEET 3) (P DYPASS I SHEET 3) POEM :75 35A IS BYPASS (SHEET 3) TRSHE :YLSE CHAMBER AESS050 SHEET (51 Cl HIGH NEUTRON FLUX ROD SOF T I ALE/K SI JONAS (5 7. MANLIA ROE H 1HDRAW).
>WvZ 5-GET 5:
+/-
P- 7 NOT AECUNOANT (SHEETS S (D:EE HL(WLR HENCE MANUAL NOTES DV PASS 1. HE BYPASS SIGNALS APE MODE UP BY MEANS CF TWO THREE-POSIT ION (NOTE 551) URNS UN A HIS HOLE. SW 1 OH I/N 4NA UTHOSSES Eli HER HC-4 1L UH NE-43L. SWI (UH I/N 45H BYPASSES EI)HEH HC-42L OR HU-44L. 1155 sO SISIAEL S. NL-35U AHO HL-SSU. ARE HERLIZED 1U 5 ACTUATE SUCH THAT A LOG:: 1 SIGNAL IS DEFINED TT BS PRESENT WHEN THE BISTABLE OUTPUT SOLTACE IS OH. NOT REDUNDANT OVERPOWER ODD STOP IBLCC/ AUTOMATIC & MANUAL ROD WITHDRAWAL) H-S (SHEETS 5 & 15) (SHEET 51
- PR STEAM SUPPLY SYSTEM 2U25 13N83 uos4*
NGT AEDUNEN NUCLEAR IN3TP. PEPU 2551 VES & ELOEKS UNITS 1 & 2 NUCLEAR ENGINEERING Dominion NORTH ANNA POWER STATION N - U W -5855 U) 33 ISH4N5I 2 3-JUN-2ØØ5 15: 15 PRIDR ID USING FDA DESIGN WDRK CHECK BMIS PDR WDPI< PENDING
SOURCE RANGE REACTOR TRIP POWER RANGE REACTOR TRIP INTERMEDIATE RANGE REACTOR TRIP II III IV POWER RANGE HIGH NEUTRON I II FLUX RATE REACTOR TRIP 4 ts ts ts 1+t s s 1+t S I NTERMED I ATE RANGE BLOCK CONTROL I /N41 K I/N42K I/N44K (NOTES 1 K 3) MANUAL MANUAL MANUAL RESET RESET RESET (NOTE 6) (NOTE 6) (NOTE 6) 1 (NOTE 5) IV 3 TRIP BYPASS I/N 3SA (NIS RACK) HIGH NEUTRON FLUX (HIGH SETPOINT) REACTOR TRIP I MANUAL TRIP (MAIN CONTROL BOARD) (SHEET 2) TO I.R. TO I.R. ROD STOP ROD STOP (SHEET 4) (SHEET 4) TO I.R. BY BY ROD STOP (SHEET 4) OTHERS RESET REACTOR TRIP REACTOR TRIP (SHEET 2) NOTES 1 THE REDUNDANT MANUAL BLOCK CONTROLS CONSIST OF TWO CONTROLS ON THE CONTROL BOARD FOR EACH RANGE, ONE FOR EACH TRAIN. NUCLEAR STEAM SUPPLY SYSTEM
- 2. I/N 33A IS IN LOGIC TRAIN A. FUNCTIONAL DIAGRAMS I/N 33B IS IN LOGIC TRAIN B. NUCLEAR IN5TRUMENT MANUAL TRIP SIGNALS
- 3. I/N 3BA IS IN LOGIC TRAIN A. Cr)
UNITS 1 I/N 38B IS IN LOGIC TRAIN B. 2 If)
- 4. I/N 47A IS IN LOGIC TRAIN A. If)
(0 I/N 47B IS IN LOGIC TRAIN B. 1? 5, TWO COMPUTER INPUTS ARE CONNECTED TO THIS CIRCUIT, INDIVIDUAL FOR EACH TRAIN.
- 6. MANUAL RESET CONTROLS CONSIST OF FOUR MOMENTARY CONTROLS IN THE CONTROL ROOM, ONE CONTROL FOR EACH INSTRUMENT CHANNEL.
z VIRGINIA POWER 01 ORIGINAL ISSUE OE5RIPT ION V-OIN I CHKII I COON I ONPR I DATE w ID NADW5655D33 NORTH ANNA POWER STATION SH 3 OF 16 I 0
EXAMINATION ANSWER KEY
<<License Class 2008-1 RO/SRO/STA Format #2>>
Q Z 57 ID: 7161 Points: 1.00
<QQ 109961 (1410)><<Given the following conditions on Unit 1:
Shutdown is in progress per 1-OP-3.7, Unit Shutdown from Mode ito Mode 5 for Refueling Power-range channel N-41 fails LOW The crew stabilized reactor power at 10%, and then compTeted 1-AP-4.3, Malfunction of Nuclear Instrumentation to place N-41 in TRIP. The following conditions exist: 0 Permissive status light P-D2, PERM PR> 10% BLK NIS LP TRIPS, is LIT o Permissive status light P-G2, P-7 PERM AT POWER TRIPS BLKD, is LIT A control power fuse blows on Power-range channel N-42. Based upon these unit conditions the reactor
<QQ 109961 (1416)><<A. will NOT trip, and the Source-range channels will NOT energize automatically when P-6 resets.
B. will trip, and the Source-range channels will energize automatically when P-6 resets. C. will trip, and the Source-range channels will NOT energize automatically when P-6 resets. D. will NOT trip, and the Source-range channels will energize automatically when P-6 resets.>> Answer: <QQ 109961(1419)><<C>> Question 57 Details Question Type: <QQ 109961(1 401)><<Multiple Choice>> Topic: <QQ 109961(1400)><<716i: ILT-SQ SYS NI>> System ID: <QQ 109961(1445)><<109961>> User ID: <QQ 109961(1404)><<7161>> Status: <QQ 109961 (1405)><<Active>> Must Appear: <QQ 109961 (1406)><<No>> Difficulty: <QQ 109961 (1407)><<0.00>> Time to Complete: <QQ 109961(1408)><<0>> Point Value: <QQ 10996 1(1441)><<1 .00>> Cross
Reference:
User Text: User Number 1: User Number 2: Comment: VA NAPS OPS Pane: 102 of 184 2009/12/03
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 24. 01 7A1 .01 024/NEW//H/3/3 .7/3.9/7/
Given the following conditions:
- Unit 1 was initially at 100% power with 1 H EDG tagged out.
- A loss of offsite power occurred, and 1J EDG locked out.
- Operators implemented 1-ECA-0.0, Loss of All AC Power.
- Power was restored to one emergency bus from the SBO Diesel.
The crew transitioned to 1-ECA-0.2, Loss of All AC Power Recovery with SI Required, Step 1. The following conditions exist:
- Containment pressure is 18 psia and slowly increasing.
- RCS pressure is 1100 psig and stable.
- CETCs are 1300°F and slowly increasing.
- RVLIS is oscillating between 64% and 68%.
Based on these conditions, which ONE of the following identifies the procedural flowpath the crew will take to mitigate this event? A. Immediately transition to 1-FR-Cl, Response to Inadequate Core Cooling; once the core cooling function is restored, return to 1-ECA-0.2, Step 1. B. Immediately transition to 1-FR-C.2, Response to Degraded Core Cooling; once the core cooling function is restored, return to 1-ECA-0.2, Step 1. C. Continue performing 1-ECA-0.2; transition to 1-FR-C.1 when directed by 1-ECA-0.2. D. Continue performing l-ECA-0.2; transition to l-FR-C.2 when directed by l-ECA-0.2.
- a. Incorrect. Plausible since the threshold for the Core Cooling status tree has been exceeded and a transition from E-0 has occurred. However ECA-0 series procedures are one of the unique procedures that depart from the typical rules of usage that would require immediate transition to the specified path. For these plant conditions a red path, Cl, (vice orange, C.2) exists (the candidate must have detailed knowledge of the CSF status trees to ascertain this) but implementation of the FR is not permitted until specifically directed by ECA-0.2.
- b. Incorrect. The specified path (C.2) is incorrect as noted above; also implementation is not allowed until directed by ECA-0.2 as discussed.
- c. Correct. First part is correct; as discussed above CSF are not implemented until specifically directed by ECA-0.2. Second part (FR-Cl) is also correct as discussed in Distractor a.
- d. Incorrect. First part is correct as explained in Distractor a; second part is incorrect but plausible, since as mentioned if the candidate lacks detailed knowledge of the status tree they could conclude that the RVLIS indication would dictate an orange path condition (which would be true for a slightly lower CETC value).
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 In-Core Temperature Monitor (ITM) System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the fTM system controls including: (CFR: 41.5 / 45.7) Core exit temperature Tier: 2 Group: 2 Technical
Reference:
1-ECA-O.2, 1-F-O Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
p Donon NORTH ANNA POWER STATION EMERGENCY CONTINGENCY ACTION NUMBER PROCEDURE TITLE REVISION 15 1 -ECA-O.2 LOSS OF ALL AC POWER RECOVERY WITH SI REQUIRED PAGE (WITH ONE ATTACHMENT) 1 of 17 PURPOSE To provide instructions to use Engineered Safeguards Systems to recover plant conditions following restoration of AC emergency power to at least one bus. ENTRY CONDITIONS This procedure is entered from:
- 1ECA-O.O, LOSS OF ALL AC POWER, or
- 1 -ECA-O.1, LOSS OF ALL AC POWER RECOVERY WITHOUT SI REQUIRED.
CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 15 1 -ECA-O 2 LOSS OF ALL AC POWER RECOVERY WITH SI REQUIRED
. PAGE 2 of 17 j_STEP j ACTION/EXPECTEDRESPONSE I I RESPONSENOTOBTAINED I
NOTE:
- CSF Status Trees should be monitored for information only.
FRs should not be initiated before completion of Step 11.
- Setpoints in brackets []are for adverse Containment atmosphere (20 psia Containment pressure or Containment Radiation has reached or exceeded 1 .0E5 R/hr or 70% on High Range Recorder).
VERIFY BOTH TRAINS OF SI - RESET C Reset both trains of SI. 2.. CHECK RWST LEVEL GREATER THAN
- Manually align valves to establish Cold Leg 15% recirculation path for energized bus or busses:
C a) Verify Recirc Spray Sump level - GREATER THAN 8 FT 0 IN C NOT,THENGOTOStep3. b) Open Low-Head SI Pump Discharge Valves to Charging Pumps: C
- 1-SI-MOV-1863A (H Bus)
C
- 1-SI-MOV-1863B (J Bus) c) Close Low-Head SI Pump Recirc Valves:
C
- 1-SI-MOV-1885A (H Bus)
C
- 1-SI-MOV-1885C (H Bus)
C
- 1-SI-MOV-1885B (J Bus)
C
- 1-SI-MOV-1885D (J Bus)
(STEP 2 CONTINUED ON NEXT PAGE)
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- 0. 0 H
z 0 4 w -J I 0 H z I -J z 0 w 0 C-) I LU C-) cr 0 H C-) H z LU C 0 L D > LU z
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 25. 022A1 .01 025/NEW//H/3/3 .6/3.7/5/
If a loss of chilled water occurs, must be aligned to the CARFs in order to prevent containment temperature from exceeding the Technical Specification limit of A. Component Cooling Water; 105°F B. Component Cooling Water; 115°F C. Service Water; 105°F D. Service Water; 115°F
- a. Incorrect. Plausible since use of the alternate source (service water) is rarely performed, thus the candidate who does not have detailed knowledge of the system or Abnormal Procedure (AP) may conclude that since CC supplies other loads in CNTMT such as CRM shroud coolers that it would be the natural source to backup chilled water. Second part is also incorrect, but plausible since it is checked in the associated AP; however, this is an EQ concern and not the TS limit.
- b. Incorrect. First part incorrect but plausible as noted above. Second part is correct--this is the TS 3.6.5 LCO.
- c. Incorrect. First part is correct--the AP for this event will direct establishing Service Water if Chilled Water is unavailable. Second part incorrect, but plausible as discussed in Distractor a.
- d. Correct. First part is correct as discussed in Distractor c; Second part is correct-this is the TS 3.6.5 LCO.
Containment Cooling System (CCS) Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CCS controls including: (CFR: 41.5 /45.5) Containment temperature Tier: 2 Group: 1 Technical
Reference:
1-AP-35 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
0 Dorniion NORTH ANNA POWER STATION ABNORMAL PROCEDURE NUMBER PROCEDURE TITLE REVISION 18 1 -AP-35 LOSS OF CONTAINMENT AIR RECIRCULATION COOLING (WITH TWO ATTACHMENTS) PAGE 1 of 8 PURPOSE To provide instructions for responding to a loss of Containment Recirculation Cooling caused by loss of Chiller Units, Containment Ventilation Fans, or Chilled Water flow. ENTRY CONDITIONS This procedure is entered if Unit 1 is in Mode 1, 2, 3 or 4 and any of the following conditions exist:
- Increasing Containment temperature,
- Loss of either Mechanical or Steam Chiller,
- Containment Air Recirc Fan auto trip alarm or breaker disagreement indication on ventilation panel, or
- Annunciator Panel G B-2, CD TO AIR RECIRC CLRS HI-LO TEMP, is LIT CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 18 1 -AP-35 LOSS OF CONTAINMENT AIR RECIRCULATION COOLING PAGE 2 of 8 ]_STEP ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED NOTE: The loss of Containment Air Partial Pressure indicator(s) does NOT necessarily mean non-compliance with Tech Spec 3.6.4. 1._ DECLARE THE DIGITAL CONTAINMENT PARTIAL AIR PRESSURE INDICATORS IRABLE: D
- 1-LM-PI-1O1A-1 D
- 1-LM-PI-1O1B-1
- 2. INITIATE ATTACHMENT 2, CALCULATION OF CONTAINMENT AIR PARTIAL PRESSURE, WHILE CONTINUING WITH THIS PROCEDURE NOTE: Because of Environmental Qualification concerns relating to equipment life expectancy, Containment Average Air temperature should be kept less than 105 °F.
3._ CHECK CONTAINMENT AVERAGE AIR TEMPERATURE: D a) Less than 115 °F D a) Enter Action Statement of Technical Specification 3.6.5. D b) Less than 105 °F D b) Notify Engineering Department. Cf C
NUMBER PROCEDURE TITLE REVISION 18 1 -AP-35 LOSS OF CONTAINMENT AIR RECIRCULATION COOLING PAGE 3 of 8 -j_STEP_fj ACTION/EXPECTEDRESPONSE RESPONSENOTOBTAINED I 4._ VERIFY 1-CC-Fl-i 28, CHILLED WATER Do either of the following using 1 -OP-2i .1, FLOW INDICATION NORMAL CONTAINMENT VENTILATION:
- Align Chilled Water to Containment Air Recirc Fans, ,
D
- Align Service Water to Containment Air Recirc Fans.
D IF Service Water is aligned to Containment Air Recirc Fans, THEN GO TO Step 8.
- 5. CHECK 2-CD-MR-i, MECHANICAL CHILLER Do the following:
- NORMAL D a) IF the Mechanical Chiller has tripped, THEN try to restart the chiller using one o of the following:
AJk-V
- O-OP-51.5,OPERATIONOFTHE CHILLED WATER SYSTEMS:
5 / MECHANICAL CHILLER 0
- 0-OP-Si .7, RESTARTING THE
/ 0
/If Cb 11-7 MECHANICAL CHILLER FOLLOWING AN AUTOMATIC CHILLER TRIP
/
D b) IF the chiller can be started, THEN GO TO Step 8. D HNOT,THENGOTOStep7.
- 6. GOTOSTEP8
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 26. 025AK2.03 026/NEW//H/4/2 .7/2.7/4/
Given the following conditions:
- Unit 1 is on RHR, and has just entered Mode 5 for a scheduled refueling outage.
- Unit 2 is at 100% power.
- Unit 1 has one CC pump and one SW pump running.
- Unit 2 has two CC pumps and two SW pumps running.
- All Service Water spray arrays are in service.
Unit 2 experiences a catastrophic failure of the A Main Steamline inside Containment. Which ONE of the following describes the initial effect of this event on Unit 1 RHR temperature, AND includes the reason? RHR temperature will A. increase due to the decrease in total CC flow. B. increase due to the decrease in total SW flow. C. decrease due to the increase in total CC flow. D. decrease due to the increase in total SW flow.
- a. Correct. The catastrophic failure of the A Main Steamline Inside Containment will result in a CDA signal on Unit 2. A by product of the Unit 2 CDA is an auto-start of the idle SW pump and trip of the Unit 2 CC pumps along with isolation of the Unit 2 CC H!Xs. Phase B actuates, which will isolate some Unit 2 CC loads. The overriding effect of the transient is the reduction in heat transfer through the Unit 1 RHR heat exchangers to the CC system, thus resulting in an increase in temperature.
- b. Incorrect. Plausible, since as noted above several things take place as a result of the event on Unit 2.
A slight reduction in SW flow to Unit 1 is a consequence of initiation of RSHX flow on Unit 2, however this is compensated for by auto-start of the idle Unit 1 SW pump. Thus as discussed above while the change in SW system alignment is a consideration, it is not the reason for the plant response on Unit 1.
- c. Incorrect. Plausible since as discussed above isolation of CC loads on Unit 2 would tend to increase CC flow on Unit 1. If the candidate fails to consider ALL of the effects of the steam break on Unit 2 they may erroneously conclude that more heat removal in the Unit 1 RHR H/Xs is occurring.
- d. Incorrect. Plausible since as discussed above auto-start of the idle SW pump will result. This is a case again where the candidate who focuses on only a few aspects of the Unit 2 steam break vice considering the aggregate effect may draw an erroneous conclusion based on incomplete analysis of the event.
QUESTIONS REPORT for NAPS 2010 RD NRC Exam rev3 Loss of Residual Heat Removal System (RHRS) Knowledge of the interrelations between the Loss of Residual Heat Removal System and the following: (CFR 41.7 /45.7) Service water or closed cooling water pumps Tier: 1 Group: 1 Technical
Reference:
AR E-A5, EOP 2-E-O Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
VIRGINIA POWER l-EI-CB-21E ANNUNCIATOR A5 1-AR-E--A5 NORTH ANNA POWER STATION REV. 2 APPROVAL: ON FILE Effective Date:01/24/01 RHR HX > 150°F CC OUTLET HI TEMP/ > 9000 GPM HI FLOW 1.0 Probable Cause 1.1 Insufficient component cooling flow 1.2 Insufficient service water flow 1.3 High rate of primary system cooldown 1.4 Starting of second Component Cooling Water Pump with only one RHR HX in service 1.5 Failure of Controlling CC System PCV 2.0 Operator Action 2.1 Adjust CC flow to RHR HX By throttling 1-CC-MOV-100A and/or B as required. 2.2 If required, place a second RHR HX in service. 2.3 If required, increase SW flow to CC HXs. 2.4 If required, start an additional CC pump using 1-OP51.1, Component Cooling System. NOTE: Minor temperature excursions above 180°F for short periods of time will not cause piping damage, however, sustained operation above 180°F is not allowed without Engineering Analysis/Review. (Reference 3.6) 2.5 Maintain CC piping less than 180°F. IF sustained temperature of CC piping exceeds 180°F, THEN do the following: (Reference 3.5)
- a. Declare affected piping INOPERABLE
- 1CCPP18.00CC--PIPE326151Q3, Discharge Piping from 1-RH-E-1A
- 1CCPP18.00--CCPIPE-327151Q3, Discharge Piping from 1-RH-E-1B
- b. Submit P1 2.6 Maintain RCS cooldown within limits.
2.7 Verify proper CC lineup to RHR HXs. çi5 2.8 Verify proper CC pressure. 3.0 References 3.1 11715FM79A, B Component Cooling
3.2 11715ESK1OAAX 3.3 NAPS instrumentation manual (pages CC 100, 101, 110 and 111) 3.4 NAPS Setpoint Doc 3.5 P1 N20001881, CC Return Piping Max Operating Temp 180°F 3.6 ET CEMO1-0003, Rev. 0, Operating Limitation for Component Cooling Temperature from RHR Heat Exchangers. 4.0 Actuation 4.1 1CCTC149A 4.2 1CCTC149B 4.3 1CCFC132A 4.4 1CCFC-132B
NUMBER ATTACHMENT TITLE ATTACHMENT 2-E-O 4 EQUIPMENT VERIFICATION REVISION PAGE 44 3of4 -_STEP_[j ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED
*7 CHECK IEQUlRED:
a) Verify EITHER of the following conditions - D a) RETURN this Attachment to SRO MET: for continued monitoring.
- Containment pressure HAS-EXCEEDED 28 PSIA OR
- Verify BOTH of the following:
- Containment pressure - HAS EXCEEDED 20 psia AND
- Containment Radiation GREATER THAN 2 R/hr D b) ManuallyactuateQDA (Dc)VerifYCCPuTED D c) Stop CC Pumps.
EJ d) Stop all RCPs D e) Verify QS Pumps - RUNNING D e) Manually start QS Pumps. 2 f) Verify QS Pump Discharge MOVs - OPEN: C f) Manually open valves. D
- 2-QS-MOV-201A
- 2-QS-MOV-201 B (STEP 7 CONTINUED ON NEXT PAGE)
NUMBER ATTACHMENT TITLE ATTACHMENT 2-E-O 4 EQUIPMENT VERIFICATION REVISION PAGE 44 2of4 _STEP [j ACTION/EXPECTEDRESPONSE I I RESPONSENOTOBTAINED I
- 4. VERIFY SI PUMPS - RUNNING: D Manually start pumps.
- Two Charging Pumps - RUNNING AND D
- Both Low-Head SI Pumps - RUNNING
- 5. VERIFY FOUR SERVICE WATER PUMPS - D Manually start pumps.
RUNNING
/ D IF less than 4 Service Water Pumps
/ are running, THEN ensure Unit 1 ii Operator initiates O-AP-47, UNIT
\ OPERATION DURING OPPOSITE UNIT EMERGENCY.
- 6. CHECK IF MAIN STEAMLINES SHOULD I BE ISOLATED:
I
/ a) Check the following: D a) RETURN this Attachment to SRO for continued monitoring.
D Annunciator Panel D E-3 LIT
/ * - / OR D
- Containment pressure HAS
( EXCEEDED 18 PSIA D b) Verify MSTVs and Bypass Valves CLOSED - U b) Manually close valves. UJ 2 u f.) v / d /i
ç I Sc) cJ U2
NUMBER ) ATTACHMENT TITLE ATTACHMENT 4 2-E-O EX9S1N VERIFICATION OF PHASE B ISOLATION 2 PAGE 44 2of9 -j_STEP ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED VERIFY THE FOLLOWING AUTOMATIC OPERATIONS ON THE H SAFEGUARDS PANEL: (Continued) D c) Verify Sece Water - ISOLATED TO ) D Manually close valve: DC CATEXCHA/ - S 2/l vj D d) Verify Service Water ALIGNED TO RECIRC
- D Manually open valves:
SPRAY HEAT EXCHANGERS (all red lights lit or check each valve below): DAlI red lights lit. OPEN (RED) OPEN (RED) OPEN (RED) OPEN (RED) 2-SW-MOV-2O3A 2-SW-MOV-2O3D 2-SW-MOV-2O4A 2-SW-MOV-2O4D 2-SW-MOV-2O1 A 2-SW-MOV-2O1 C 2-SW-MOV-2O5A 2-SW-MOV-2O5C D e) Verify Casing Cooling - ALIGNED AND D Manually do operations as indicated: RUNNING: RUNNING (RED) OPEN (RED) OPEN (RED) 2-RS-P-3A 2-RS-MOV-2OOA 2-RS-MOV-201 B
NUMBERN ATTACHMENT TITLE ATTACHMENT
,z2-E-O ) 2 , VERIFICATION OF PHASE B ISOLATION
( REVISION PAGE 44- 5 of 9 _STEP [IACTION/EXPECTED
RESPONSE
I RESPONSE NOT OBTAINED
- 4. VERIFY THE FOLLOWING AUTOMATIC OPERATIONS ON THE J SAFEGUARDS PANEL:
7 f y Service Water ISOLATED TO
- D Manually close valve:
/ CC HEAT EXCHANGERS:
,,&LED (GREEN)
( 2-SW-MOV-2O8B E1 d) Verify Service Water ALIGNED TO RECIRC
- D Manually open valves:
SPRAY HEAT EXCHANGERS (all red lights lit or check each valve below): IAll red lights lit. OPEN (RED) OPEN (RED) OPEN (RED) OPEN (RED) _2-SW-MOV-203B _2-SW-MOV-203C 2-SW-MOV-2O4B _2-SW-MOV-204C _2-SW-MOV-201 B 2-SW-MOV-201 D _2-SW-MOV-205B _2-SW-MOV-205D C e) Verify Casing Cooling - ALIGNED AND C Manually do operations as indicated: RUNNING: RUNNING (RED) OPEN (RED) OPEN (RED) 2-RS-P-3B 2-RS-MOV-200B A
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 27. 026A2.05 027/NEW//H13/3.7/4.1/5/
Given the following conditions:
- Unit 1 was initially at 100% power with Quench Spray pump 1-QS-P-1B tagged out.
- A LOCA occurred 7 minutes ago.
The crew is performing 1-E-0, Reactor Trip or Safety Injection, Attachment 2, Verification of Phase B Isolation. The crew notes the following indications:
- 1-QS-MOV-102A, Chemical Addition Tank Outlet Valve, GREEN and RED lights both OFF.
- 1 -QS-MOV-1 02B, Chemical Addition Tank Outlet Valve, GREEN light ON, RED light OFF.
- 1-QS-Ll-101, Chemical Addition Tank Level, 90% and stable.
Which ONE of the following describes the action required by 1-E-0, Attachment 2, AND includes the consequence if the above malfunction is not corrected? A. Dispatch operator to locally open 1 -QS-MOV-1 02A, Chemical Addition Tank Outlet Valve; reduced boron concentration of containment sump water. B. Dispatch operator to locally open 1-QS-MOV-102A, Chemical Addition Tank Outlet Valve; reduced effectiveness of containment Iodine removal. C. Manually open 1-QS-MOV-102B, Chemical Addition Tank Outlet Valve; reduced boron concentration of containment sump water. D. Manually open 1-QS-MOV-102B, Chemical Addition Tank Outlet Valve; reduced effectiveness of containment Iodine removal.
- a. Incorrect. Plausible since the operator may erroneously assume a need to have the A MOV open since only the A QS pump is available. Second part incorrect but plausible, since the candidate who has only cursory knowledge would likely believe that the CAT, like other ECCS tanks, contains a solution of boric acid and erroneously reason that this condition would result if it is not injected.
- b. Incorrect. First part incorrect but plausible as noted above; Second part is correct as described in the TS bases, one of the purposes is to maximize the effectiveness of the Iodine removal mechanism, however the candidate who lacks detailed knowledge of the system/TS Bases may focus on the function of minimizing corrosion and thus overlook this aspect.
- c. Incorrect. First part is correct lAW 1-E-0, att. 2 the operator is directed to manually open valves, based on the conditions given this action will satisfy the safety function and thus not warrant any additional local action. Second part incorrect but plausible as discussed in Distractor a.
- d. Correct. First part is correct as discussed in Distractor c. Second part is also correct as explained in distractor b.
QUESTIONS REPORT for NAPS 2010 RD NRC Exam rev3 Containment Spray System (CSS) Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5/43.5/45.3/45.13) Failure of chemical addition tanks to inject Tier: 2 Group: 1 Technical
Reference:
EOP E-O ATT.2, TS 3.6.8 BASES Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
NUMBER ATTACHMENT TITLE ATTACHMENT 1-E-O 2 VERIFICATION OF REVISION PHASE B ISOLATION PAGE 42 lof9 H_STEP ACTION/EXPECTEDRESPONSE RESPONSENOTOBTAINED I
- 1. VERIFY THE FOLLOWING AUTOMATIC OPERATIONS ON THE H SAFEGUARDS PAN EL:
D a) Verify Phase B isolation valves (located on a Manually close valves: the lower right corner of the panel) CLOSED (all green lights lit or check each valve below): aAll green lights lit. TRIP VALVES (TV) CLOSED (GREEN) CC- CC- CC- CC _105A .1O2A .104A-1 1O1A CC- CC- CC- CC 105B 102C 1O4B-1 103A CC- CC- CC- IA 105C _102E 1O4C-1 _102A NOTE: 1 -QS-MOV-1 02A, CHEMICAL ADDITION TANK A OUTLET VALVE, opens following a 5-minute time delay. a b) Verify Quench Spray ALIGNED AND
- a Manually do operations as indicated:
RUNNING: C CL 0 RUNNING (RED) OPEN (RED) 1-QS-P-1A 1-QS-MOV-1O1A 1 -QS-MOV-1 OOA 1-QS-MOV-102A I? (STEP 1 CONTINUED ON NEXT PAGE)
NUMBER ATTACHMENT TITLE ATTACHMENT 1-E-O 2 VERIFICATION OF PHASE B ISOLATION REVISION PAGE 42 2of9 H_STEP ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED
- 1. VERIFY THE FOLLOWING AUTOMATIC OPERATIONS ON THE H SAFEGUARDS PANEL: (Continued)
C c) Verify Service Water ISOLATED TO
- C Manually close valve:
CC HEAT EXCHANGERS: CLOSED (GREEN) 1 -SW-MOV-1 08A C d) Verify Service Water ALIGNED TO RECIRC
- C Manually open valves:
SPRAY HEAT EXCHANGERS (all red lights lit or check each valve below): CAll red lights lit. OPEN (RED) OPEN (RED) OPEN (RED) OPEN (RED) 1 -SW-MOV-1 03A 1 -SW-MOV-1 03D 1 -SW-MOV-1 04A _1 -SW-MOV-1 04D 1 -SW-MOV-1 01 A 1 -SW-MOV-1 01 C 1 -SW-MOV-1 05A 1 -SW-MOV-1 05C C e) Verify Casing Cooling - ALIGNED AND C Manually do operations as indicated: RUNNING: RUNNING (RED) OPEN (RED) OPEN (RED) 1 -RS-P-3A 1 -RS-MOV-1 OOA 1 -RS-MOV-1 01 B
NUMBER ATTACHMENT TITLE ATTACHMENT 1-E-O 2 VERIFICATION OF PHASE B ISOLATION REVISION PAGE 42 3of9 ACTION/EXPECTEDRESPONSE RESPONSENOTOBTAINED H_STEP_[I
- 2. STOP 1-IA-C-2A, A CONTAINMENT INSTRUMENT AIR COMPRESSOR, BY PRESSING BOTH OFF BUTTONS
- 3. CHECK H SAFEGUARDS PANEL -
RECIRC SPRAY STATUS: D a) Verify Recirc Spray ALIGNED:
- D a) Manually align valves:
OPEN (RED) 1 -RS-MOV-1 56A 1 -RS-MOV-1 55A EJ b) Verify RWST level - LESS THAN 60% D b) WHEN RWST is less than 60%, THEN perform Step 3.c. EJ Continue with Step 4. EJ c) Verify Recirc Spray Pumps - RUNNING: c) Manually do operations as indicated: RUNNING (RED) RUNNING (RED) _1 -RS-P-1 A 1 -RS-P-2A (2 minute time delay after RWST Level <60%) (When RWST Level <60%)
NUMBER ATTACHMENT TITLE ATTACHMENT 1-E-O 2 VERIFICATION OF PHASE B ISOLATION REVISION PAGE 42 4of9 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED j_STEP_[J 4._ VERIFY THE FOLLOWING AUTOMATIC OPERATIONS ON THE J SAFEGUARDS PAN EL: C a) Verify Phase B isolation valves (located on C Manually close valves: the lower right corner of the panel) CLOSED (all green lights lit or check each valve below): CAll green lights lit. TRIP VALVES (TV) CLOSED (GREEN) CC- CC- CC- CC 100A 102B 104A-2 1O1B CC- CC- CC- CC 100B 102D 1048-2 1O3B CC- CC- CC- IA 1OOC _102F 104C-2 102B NOTE: 1 -QS-MOV-1 02B, CHEMICAL ADDITION TANK B OUTLET VALVE, opens following a 5-minute time delay. C b) Verify Quench Spray ALIGNED AND
- C Manually do operations as indicated:
RUNNING: V RUNNING (RED) OPEN (RED) 1 -QS-P-1 B 1 -QS-MOV-1 01 B 1 -QS-MOV-1 OOB 1 -QS-MOV-1 02B (STEP 4 CONTINUED ON NEXT PAGE)
NUMBER ATTACHMENT TITLE ATTACHMENT 1-E-O 2 VERIFICATION OF PHASE B ISOLATION REVISION PAGE 42 5of9 J_STEP_[-H ACTION/EXPECTED RESPONSE I RESPONSENOTOBTAINED I
- 4. VERIFY THE FOLLOWING AUTOMATIC OPERATIONS ON THE J SAFEGUARDS PANEL: (Continued)
D c) Verify Service Water ISOLATED TO
- D Manually close valve:
CC HEAT EXCHANGERS: CLOSED (GREEN) 1 -SW-MOV-1 08B D d) Verify Service Water ALIGNED TO RECIRC
- D Manually open valves:
SPRAY HEAT EXCHANGERS (all red lights lit or check each valve below): DAlI red lights lit. OPEN (RED) OPEN (RED) OPEN (RED) OPEN (RED) _1 -SW-MOV-1 03B _1 -SW-MOV-1 03C _1 -SW-MOV-1 04B _1 -SW-MOV-1 04C _1 -SW-MOV-1 01 B _1 -SW-MOV-1 01 D _1 -SW-MOV-1 05B _1 -SW-MOV-1 05D D e) Verify Casing Cooling ALIGNED AND
- D Manually do operations as indicated:
RUNNING: RUNNING (RED) OPEN (RED) OPEN (RED) 1 -RS-P-3B 1 -RS-MOV-1 OOB 1 -RS-MOV-1 01 A
NUMBER ATTACHMENT TITLE ATTACHMENT 1-E-0 2 VERIFICATION OF PHASE B ISOLATION REVISION PAGE 42 6of9 -_STEP ACTION/EXPECTED RESPONSE I RESPONSE NOTOBTAINED I
- 5. CHECK J SAFEGUARDS PANEL -
RECIRC SPRAY STATUS: D a) Verify Recirc Spray ALIGNED:
- D a) Manually align valves:
OPEN (RED) 1 -RS-MOV-1 56B 1 -RS-MOV-1 55B D b) Verify RWST level - LESS THAN 60% D b) WHEN RWST is less than 60%, THEN pertorm Step 5.c. EJ Continue with Step 6. D c) Verify Recirc Spray Pumps - RUNNING: D c) Manually do operations as indicated: RUNNING (RED) RUNNING (RED) 1 -RS-P-1 B 1 -RS-P-2B (2 minute time delay after RWST Level <60%) (When RWST Level <60%)
NUMBER ATTACHMENT TITLE ATTACHMENT 1-E-O 2 VERIFICATION OF PHASE B ISOLATION REVISION PAGE 42 7of9 H_STEP ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED
- 6. VERIFY THE FOLLOWING AUTOMATIC J Manually do operations as indicated:
OPERATIONS ON THE UNIT 1 VENTILATION PAN EL: D a) Verify all Containment Air Recirc Fans - STOPPED: 1 -H V-F-lA 1 -HV-F-1 C 1 -HV-F-1 B -HV-F-1 C D b) Verify all CRDM fans - STOPPED: _1-HV-F-37B 1-HV-F-37D D c) Verify Service Water Supply and Return for Recirc Air Coolers CLOSED (GREEN): CLOSED (GREEN) CLOSED (GREEN) 1 -SW-TV-i 01 A-i 1 -SW-TV-i 01 A-2 1 -SW-TV-i 01 B-i -SW-TV-i 01 B-2 (STEP 6 CONTINUED ON NEXT PAGE)
NUMBER ATTACHMENT TITLE ATTACHMENT 1-E-O 2 VERIFICATION OF PHASE B ISOLATION REVISION PAGE 42 8of9 STEP ACTION/EXPECTEDRESPONSE I I RESPONSENOTOBTAINED I
- 6. VERIFY THE FOLLOWING AUTOMATIC OPERATIONS ON THE UNIT 1 VENTILATION PANEL: (Continued)
D d) Verify Iodine Filter Banks IN SERVICE FOR SAFEGUARDS VENTILATION: FILTER (RED) FILTER (RED) (H Train) (J Train) 1-HV-AOD-1O7A-1, 2, 3, & 4 1-HV-AOD-1O7B-1, 2,3, & 4 NOTE: The sample pumps automatically start following a 2-minute time delay. The Low Flow Alarm is enabled after an additional 30 seconds.
- 7. VERIFY THE FOLLOWING SAMPLE PUMP J Notify SRO.
RED LOW FLOW LIGHTS NOT LIT ON THE UNIT 1 RADIATION MONITORING PANEL: NOT LIT NOT LIT NOT LIT NOT LIT
-SW-P-5 1 -SW-P-8 1 -SW-P-6 1 -SW-P-7
H z 0 w 0)
- UJ
< Q0) H cl) > G) HW E 0) WCJ) C G) W (ES E Wc (ES (ES WQC CESO WE
- 0) _C C Ci 5-. (ES 0
- 0 0 WZ O.
z ES) Cl) (ES (ES 0 OI U)d H 0 0 0
-J 0
w C/) D H W H z 0 0 I - 0 0 cx: z 00 UJH 0 0 U H 0< w WI
> U)Q Cd) 0 df)
>0w L65H z wW<
cr >U) w C 0 U) C-j D > z 1 w C Cr
NUCLEAR DESIGN INFORMATION PORTAL Chemical Addition System B 3.6.8 B 3.6 CONTAINMENT SYSTEMS B 3.6.8 Chemical Addition System BASES BACKGROUND The Chemical Addition System is a subsystem of the Quench Spray System that assists in reducing the iodine fission product inventory in the containment atmosphere resulting from a Design Basis Accident (DBA). Radioiodine in its various forms is the fission product of primary concern in the evaluation of a DBA. It is absorbed by the spray from the containment atmosphere. To enhance the iodine absorption capacity of the spray, the spray solution is adjusted to an alkaline pH that promotes iodine hydrolysis, in which iodine is converted to nonvolatile forms. Because of its stability when exposed to radiation and elevated temperature, sodium hydroxide (NaOH) is the preferred spray additive. The NaOH added to the spray also ensures a pH value of between 7.0 and 8.5 of the solution recirculated from the containment sump. This pH band minimizes the evolution of iodine as well as the occurrence of chloride and caustic stress corrosion on mechanical systems and components. The Chemical Addition System consists of one chemical addition tank, two parallel redundant motor operated valves in the line between the chemical addition tank and the refueling water storage tank (RWST), instrumentation, and a recirculation pump. The NaOH solution is added to the spray water by a balanced gravity feed from the chemical addition tank through the connecting piping into a weir within the RWST. There, it mixes with the borated water flowing to the spray pump suction. Because of the hydrostatic balance between the two tanks, the flow rate of the NaOH is controlled by the volume per foot of height ratio of the two tanks. This ensures a spray mixture pH that is 8.5 and 10.5. The Quench Spray System actuation signal opens the valves from the chemical addition tank to the spray pump suctions or the quench spray pump start signal opens the valves from the chemical addition tank after a 5 minute delay. The 12% to 13% NaOH solution is drawn into the spray pump suctions. The chemical addition tank capacity provides for the addition of NaOH solution to all of the water sprayed from the RWST into containment. The percent solution and volume of solution (continued) North Anna Units 1 and 2 B 3.6.8-1 Revision 36
NUCLEAR DESIGN INFORMATION PORTAL Chemical Addition System B 3.6.8 BASES BACKGROUND sprayed into containment ensures a long term containment (continued) sump pH of 7.0 and 8.5. This ensures the continued iodine retention effectiveness of the sump water during the recirculation phase of spray operation and also minimizes the occurrence of chloride induced stress corrosion cracking of the stainless steel recirculation piping. Maintaining the sump fluid pH less than or equal to 8.5 ensures that there is adequate NPSH available to the ECCS and RSS pumps with post LOCA debris and chemical precipitant loading on the containment sump strainer. APPLICABLE The Chemical Addition System is essential to the removal of SAFETY ANALYSES airborne iodine within containment following a DBA. Following the assumed release of radioactive materials into containment, the containment is assumed to leak at its analysis value volume following the accident. The plant accident dose calculations use an effective containment coverage of 70% of the containment volume. The containment safety analyses implicitly assume that the containment atmosphere is so turbulent following an accidental release of high energy fluids inside containment that, for heat removal purposes, the containment volume is effectively completely covered by spray. The DBA response time assumed for the Chemical Addition System is based on the Chemical Addition System isolation valves beginning to open 5 minutes after a QS pump start. The DBA analyses assume that one train of the Quench Spray System is inoperable and that the entire chemical addition tank volume is added through the remaining Quench Spray System flow path. The Chemical Addition System satisfies Criterion 3 of 10 CFR 50.36(c) (2) (ii). LCO The Chemical Addition System is necessary to reduce the release of radioactive material to the environment in the event of a DBA. To be considered OPERABLE, the volume and concentration of the chemical addition solution must be sufficient to provide NaOH injection into the spray flow until the Quench Spray System has completed pumping water from the RWST to the containment sump, and to raise the (conti nued) North Anna Units 1 and 2 B 3.6.8-2 Revision 36
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 28. 026AA1.03 028/NEW//L/2/3.6/3.6/8/
In the event that CC is lost and cannot be readily restored, 1-AP-15, Loss of Component Cooling, will direct the operator to align to the Fuel Pit Coolers. A. Chilled Water B. Fire Protection Water C. Primary Grade Water D. Service Water
- a. Incorrect. Plausible since this system is in the aux bldg and the operator who doesnt have detailed knowledge of the procedure and systems & interconnections might select it.
- b. Incorrect. Plausible since this is a makeup source to the Fuel Pit, and the candidate may confuse or erroneously assume that it could be a backup cooling source also.
- c. Incorrect. Plausible since this is a backup cooling source to some loads (e.g. Charging Pumps) and as noted above the operator who doesnt have detailed knowledge of the procedure and systems &
interconnections might select it.
- d. Correct. 1-AP-15, Attachment 3 provides this direction.
Loss of Component Cooling Water (CCW) Ability to operate and I or monitor the following as they apply to the Loss of Component Cooling Water: (CFR 41.7 /45.5 I 45.6) SWS as a backup to the COWS Tier: 1 Group: 1 Technical
Reference:
1-AP-15 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
NUMBER PROCEDURE TITLE REVISION 21 1-AP-15 LOSS OF COMPONENT COOLING PAGE 7 of 9 H_STEP ACTION/EXPECTED RESPONSE ( RESPONSE NOT OBTAINED 17._ DETERMINE IF CC SHOULD BE GO TO Step 19. CROSS-TIED: 0
- Check CC System - INTACT AND 0
- Check Unit 2 CC System -
AVAILABLE
- 18. USE 1-OP-51.1, COMPONENT COOLING SYSTEM, TO CROSS-TIE CC SYSTEMS 19._ REVIEW TECH SPECS FOR APPLICABILITY:
0
- TechSpec3.7.19 0
- TRM3.7.15
- 20. CHECK RHR SYSTEM - SECURED 0 CC has NOT been restored to RHR, THEN initiate 1-AP-11, LOSS OF RHR.
21._ STABILIZE THE RCS: 0 a) Stabilize RCS temperature using Condenser Steam Dump or SG PORVs 0 b) Maintain PRZR level at 80% or less
- 22. VERIFY CC FLOW TO COMMON 0 Transfer common loads to the unaffected unit using LOADS NORMAL 1-OP-51.1, COMPONENT COOLING SYSTEM.
O IF CC has been lost to both units, THEN initiate ATTACHMENT 3.
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-15 3 SECURING COMMON CC LOADS REVISION PAGE 21 4of4
- 11. Monitor 1-FC-Tl-i 03, Spent Fuel Pit Temperature. WHEN directed by the Operating Supervisor, THEN locally align Service Water to Fuel Pit Coolers:
a) Close Fuel Pit Cooler CC Valves:
- 1-CC-273, Inlet to Fuel Building Header from CC System (AB 244 overhead, north of 2-CC-E-1 B)
- 1-CC-308, Return from Fuel Building Header to CC System (AB 244 overhead, chain valve, north of 2-CC-P-i B) b) Drain header by removing caps and opening the following drain valves (located in the Demin Alley, U-2 side, AB 244):
- 1-CC-666, Supply Drain to Fuel Building
- 1-CC-667, Return Drain from Fuel Building c) Close 1-CC-666, Supply Drain to Fuel Building, and install cap.
d) Energize and open one of the following:
- 1-SW-MOV-li3B (1-EE-BKR-lHi-2N N2), No.1 Service Water Supply Header
- 2-SW-MOV-213B (2-EE-BKR-2Hi-2N M4), No. 2 Service Water Supply Header e) Open 1-CC-272, Inlet to Fuel Building Header from SW System (AB 244 overhead, North of 2-CC-E-1 B).
f) Flush header several minutes, then close 1-CC-667 and install cap. g) Energize and open one of the following:
- 2-SW-MQV-213A (2-EE-BKR-2H1-2N M3), No. 4 Service Water Return Header
- 1-SW-MOV-113A (1-EE-BKR-1H1-2N Ni), No.3 Service Water Return Header h) Open I -CC-309, Return from Fuel Building Header to SW System (AB 244 overhead, NW of 2-CC-P-i B).
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 29. 027AK3 .01 029/NEW//H14/3 .5/3.8/3/
Given the following conditions:
- Unit 1 is at 30% power and holding for chemistry.
- The block valve for 1-RC-PCV-1455C, PRZR PORV, is closed.
1-RC-PC-1444J, PRZR Master Pressure Controller, fails to 100% demand and will not swap to MANUAL. In response to this event, the crew will A. manually open spray valves to prevent cycling a PRZR PORV. B. manually open spray valves to prevent a reactor trip. C. manually close spray valves to prevent a turbine runback. D. manually close spray valves to prevent a reactor trip.
- a. Incorrect. Plausible since the master controller is inverse acting; the candidate who is unsure of, or confuses the operation of the controller may erroneously assume that pressure would be increasing and select this distractor (the opposite train PORV is in fact functional).
- b. Incorrect. Plausible as noted above; for this case the candidate may confuse which PORV is associated with the controller, and with one PORV isolated may erroneously conclude that neither PORV will auto open at 2335 psig and thereby select this distractor.
- c. Incorrect. Plausible since pressure will decrease, and pressure is an input to the OT Delta-T turbine runback, however at this power level the reactor trip on low PRZR pressure will occur before the OTDT runback setpoint is reached.
- d. Correct. As discussed in Distractor c pressure will decrease because the failure caused spray valves to open and control group heaters will go to minimum gate; spray valves must be closed manually to prevent continue depressurization resulting from the reduced heater input.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Pressurizer Pressure Control System (PZR PCS) Malfunction Knowledge of the reasons for the following responses as they apply to the Pressurizer Pressure Control Malfunctions: (CFR 41.5,41.10 /45.6 I 45.13) Isolation of PZR spray following loss of PZR heaters Tier: 1 Group: 1 Technical
Reference:
1-AP-44, AR5 B-E6, D-B2, D-B1, B-B3, TR 4.3 TBL 4.3-1 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
Dottnen NORTH ANNA POWER STATION ABNORMAL PROCEDURE NUMBER PROCEDURE TITLE REVISION 19 1 -AP-44 LOSS OF REACTOR COOLANT SYSTEM PRESSURE (WITH ONE ATTACHMENT) PAGE 1 of 5 PURPOSE To provide operator guidance in the event of a decreasing Pressurizer pressure, but not necessarily a decreasing Pressurizer level ENTRY CONDITIONS This procedure is entered when the following conditions exists:
- Annunciator Response B Panel F-7, PRZR HIGH-LOW PRESSURE
- Annunciator Response C Panel D-1, PRZR SAFETY VALVE OR PORV OPEN
- Reactor Coolant System Pressure less than 2335 psig and PZR PORV open CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 19 1 -AP-44 LOSS OF REACTOR COOLANT SYSTEM PRESSURE PAGE 2 of 5 -j_STEP H ACTION/EXPECTEDRESPONSE I RESPONSENOTOBTAINED CAUTION: 9 any time during this procedure, then 1-E-0, REACTOR If RCS pressure is less than 1870 psi TRIP OR SAFETY INJECTION, must be initiated while continuing with this procedure. 1] CHECK PRZR PORVs CLOSED:- D Close the PORVs. D
- 1 -RC-PCV-1 455C C IF any PORV cannot be closed, THEN manually close the associated Block Valve.
C
- 1-RC-PCV-1456 4..
C H any PORV is open AND the associated Block Valve will not close, THEN GO TO 1-E-0, REACTOR TRIP OR SAFETY INJECTION, while continuing with this procedure. 2 ] CHECK MASTER PRESSURE C Put the controller in MANUAL and adjust as CONTROLLER CONTROLLING required to stabilize and restore pressure. PROPERLY
NUMBER PROCEDURE TITLE REVISION 19 1 -AP-44 LOSS OF REACTOR COOLANT SYSTEM PRESSURE PAGE 3 of 5 STEP ACTION/EXPECTEDRESPONSE I I RESPONSENOTOBTAINED I 3 ] CHECK PRZR SPRAY VALVES - D Manually close valves using controller. CLOSED: D
- 1 -RC-PCV-1 455A D Verify PRZR spray valves closed.
- 1RCPCV1455B IF NOT, THEN place failed valve remote close switch in CLOSE:
D
- 1-RC-SOV-1455A, 1-RC-PCV-1455A REMOTE CLOSE SOV 1i
- 1-RC-SOV-1455B, 1-RC-PCV-1455B REMOTE CLOSE SOV IF valves can NOT be closed, THEN do the following:
D a) GO TO 1-E-O, REACTOR TRIP OR SAFETY INJECTION, while continuing with this procedure. l b) Stop 1-RC-P-1C.
- 4. VERIFY ALL PRZR HEATERS- D Manually energize PRZR heaters as required to ENERGIZED maintain desired PRZR pressure.
- 5. CHECK 1 -CH-HCV-i 311, AUXILIARY SPRAY VALVE CLOSED
VIRGINIA POWER 1-EI-CB-21B ANNUNCIATOR E6 1-AR--B-E6 NORTH ANNA POWER STATION REV. 0 APPROVAL: ON FILE Effective Date:11/08/96 PRZ 1RCPC1444J PRESS CONT 92.5% Controller HI OUTPUT Output NOTE: Pressurizer PORV 1-RCPCV1455C opens coincident with this alarm. Both Pressurizer Spray valves may also be open. 1.0 Probable Cause 1.1 PZR master pressure controller failed Hi 1.2 Pressure channel 1RCPC1444 failed Hi 2.0 Operator Action 2.1 IF RCS is NOT in an overpressure condition, THEN go to lAP44, Loss of Reactor Coolant System Pressure 2.2 IF the RCS is in an overpressure condition, THEN reduce RCS pressure to normal using heaters and spray and ensure that 1-RCPCV-1455C closes. 3.0 References 3.1 11715ESK1OAAH 3.2 W drawing 5655D33 3.3 PLS Document 3.4 Unit 1 Loop Book, page RC110 4.0 Actuation 4.1 Pressure comparator 1RCPC1444E
VIRGINIA POWER 1-EI-CB-21D ANNUNCIATOR B2 1-AR-D-B2 NORTH ANNA POWER STATION REV. 1 SNSOC APPROVAL: ON FILE Effective Date:05/06/02 PRESSURI ZER 1870 psig on LO PRESS 2/3 Channels RX TRIP Interlocked with P7 1.0 Probable Cause 1.1 Loss of Pressurizer Heaters 1.2 Loss of Reactor Coolant 1.3 Excessive spray-down 1.4 Excessive cooldown 1.5 Failure of 1RCPC1444J, Pressurizer Master Pressure Controller 2.0 Operator Action 2.1 IF the Reactor is tripped, THEN GO TO 1E-0, Reactor Trip or Safety Injection. 2.2 IF the Reactor is NOT tripped, THEN verify RCS pressure is normal by using alternate and redundant indications. 2.3 IF RCS pressure is NOT normal OR unable to determine that the Reactor is in a safe operating condition, THEN trip the Reactor and GO TO 1E0, Reactor Trip or Safety Injection. 2.4 IF Reactor power is < 10% AND P-7 PERM AT POWER TRIPS BLKD (Panel P-G2) is NOT LIT, THEN begin an orderly shutdown and initiate Actions in accordance with Tech Spec 3.0.3 (ITS 3.0.3) for inoperable Solid State Protection System. 2.5 IF Reactor power is > 10%, RCS pressure is > 1870 psig, AND the logic of Step 4.1 exists, THEN begin an orderly shutdown and initiate Actions in accordance with Tech Spec 3.0.3 (ITS 3.0.3) for inoperable Solid State Protection System. 2.6 IF all plant parameters are normal AND the alarm is due to unknown instrument malfunctions, THEN immediately contact Instrument Department to determine cause of alarm. 3.0 References 3.1 Westinghouse Logic NADW5655D33 3.2 11715FM 93B 3.3 11715ESK1OD, 1OAAA 3.4 Instrument Loops 11715RC070, 072, 074
VIRGINIA POWER 1-EI-CB-21D ANNUNCIATOR Bi 1-AR-D-B1 NORTH ANNA POWER STATION REV. 2 SNSOC APPROVAL: ON FILE Effective Date:05/06/02 PRESSURI ZER HI PRESS 2360 psig on RX TRIP 2/3 Channels 1.0 Probable Cause 4S 1.1 Excessive loss of load 1 V IS
\.
1.2 Loss of spray ) 1.3 Pressurizer heater failure 1.4 Failure of 1RCPC1444J, Pressurizer Master Pressure Controller 2.0 Operator Action 2.1 IF the Reactor is tripped, THEN GO TO 1E-0, Reactor Trip or Safety Injection. 2.2 IF the Reactor is NOT tripped, THEN verify RCS pressure is normal by using alternate and redundant indications. 2.3 IF RCS pressure is NOT normal OR unable to determine that the Reactor is in a safe operating condition, THEN trip the Reactor and GO TO 1E-0, Reactor Trip or Safety Injection. 2.4 IF RCS pressure is normal AND at least 2/3 PRZR HI PRESS CHNL I / II / III (Panel MHl, H2, H3) status alarms are LIT, THEN begin an orderly shutdown and initiate Actions in accordance with Tech Spec 3.0.3 (ITS 3.0.3) for inoperable Solid State Protection System. 2.5 IF all plant parameters are normal AND the alarm is due to unknown instrument malfunctions, THEN immediately contact Instrument Department to determine cause of alarm. 3.0 References 3.1 Westinghouse Logic NA-DW5655D33 3.2 117 15FM93B 3.3 11715ESK1OD, 1OAAA 3.4 Instrument Loops 117l5RC070, 072, 074 3.5 Tech Spec 3.0.3 (ITS 3.0.3) 3.6 1E0, Reactor Trip or Safety Injection 3.7 T.S. Change 330, Amendment 200, Pressurizer Safety Valve Lift Setpoints tolerance change 4.0 Actuation
VIRGINIA POWER l-EI-CB-2lB ANNUNCIATOR B3 1-AR B B 3 NORTH ANNA POWER STATION REV. 1 APPROVAL: ON FILE Effective lDate:09/10/08 OVERTEMP T ROD STOP TURB RUNBACK .- Variable CH I-Il-Ill 1.0 Probable Cause c 1.1 Uncontrolled RCCA bank withdrawal at power 1.2 Uncontrolled boron dilution 1.3 Decreasing pressurizer pressure 1.4 Xenon oscillation 1.5 Excessive load increase 1.6 Channel in test 2.0 Operator Action 2.1 IF an approved test in progress, THEN verify normal indications and alarm cleared at the completion of testing 2.2 IF overtemp zT condition exists, THEN verify auto runback or reduce Tavg and turbine load using l-AP-38, Excessive Load Increase. 2.3 IF failed channel, THEN GO TO l-AP-3, Loss of Vital Instrumentation. 3.0 References 3.1 ll7ls-FM-93A 3.2 westinghouse Logic 5655D33 3.3 ll715-ESK-1OAAP 3.4 Unit 1 Loop Book, page RC-lOO 4.0 Actuation 4.1 l-RC-TC-l4l2C, l-RC-TC-l422C or l-RC-TC--l432C
NUCLEAR DESIGN INFORMATION PORTAL RTS Instrumentation Trip Setpoints
) LC\
Table 4.3-1 (page 3 of 4) Reactor Trip Instrumentation Trip Setpoints 7Soi Note 1: S+ UVU JC9 Overtemperature AT 1 ATO[K K(P P) Q bC fi(AI)1
+/-
_ jAJ c tUflA_) evJ S ( Where: AT 0 = Indicated AT at RTP T = Average temperature, °F Lt-T = Indicated Tavg at RTP 586.8°F P = Pressurizer pressure, psig 1 P = 2235 psig (indicated RCS nominal operating pressure) ,& C 1 +/--c S 1
= The function generated by the lead-lag controller for 1 +T S
2 dynamic compensation 1 &
-c 2 t = Time constants utilized in the lead-lag controller for Tavg
= 25 secs, 2 = 4 secs T
S = Lapl ace transform operator (sec) K 1 = 1.264 2 K = 0.0220 3 K = 0.001152 and f(AI) is a function of the indicated difference between top and bottom 1 detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that: (i) for q - q between -35 percent and +3 percent, f (AI) = 0 (where qt 1 and q are percent RTP in the top and bottom halves of the core respectively, and q + q is total THERMAL POWER in percent of RTP). (ii) for each percent that the magnitude of (cit - qb) exceeds -35 percent, the AT trip setpoint shall be automatically reduced by 1.67 percent of its value at RTP. (iii) for each percent that the magnitude of (cit - qb) exceeds +3 percent, the AT trip setpoint shall be automatically reduced by 2.00 percent of its value at RTP. NAPS TRM 4.3-4 Rev 46, 04/02/04
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 30. 029K1 .03 03 0/BANK//L/3/3 .6/3.8/8/
Containment Purge is in service on Unit 1. Which ONE of the following identifies conditions that will automatically isolate Containment Purge? A. 1-RM-RMS-159, Containment Particulate Radiation Monitor, Hi-Hi alarm. 1-VG-RI-179-1, MGP Vent Stack A Rad monitor, Hi alarm. B. 1-RM-RMS-162, Manipulator Crane Radiation Monitor, Hi-Hi alarm. 1-VG-Rl-180-1, MGP Vent Stack B Rad monitor, Hi alarm. C. 1-RM-RMS-159, Containment Particulate Radiation Monitor, Hi-Hi alarm. 1-RM-RMS-162, Manipulator Crane Radiation Monitor, Hi-Hi alarm. D. 1-RM-RMS-160, Containment Gaseous Radiation Monitor, Hi-Hi alarm. 1-VG-Rl-180-1, MGP Vent Stack B Rad monitor, Hi alarm.
- a. Incorrect. First one is correct. Second one is incorrect but plausible since this is a direct pathway from the containment to atmosphere and the candidate who lacks detailed systems knowledge might conclude that this would be logical (since if you had an alarm on a monitor that is measuring release directly to the environment, you would want an automatic feature for stopping the release).
- b. Incorrect. First one is correct. Second part is incorrect but plausible since again the candidate who lacks detailed systems knowledge might conclude that it would be logical for the purge system to isolate and default to this distractor. It should also be noted that one of the sister MGPs for process vents does cause automatic isolation of CNTMT vacuum TVs and this function could easily be confused with the CNTMT Purge System.
- c. Correct. These are both correct.
- d. Correct. First one is correct. Second part is is incorrect but plausible as discussed above.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Containment Purge System (CPS) Knowledge of the physical connections and/or cause effect relationships between the Containment Purge System and the following systems: (CFR: 41.2 to 41.9 /45.7 to 45.8) Engineered safeguards Tier: 2 Group: 2 Technical
Reference:
1-AP-5, O-AP-5.2 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info: NAPS does not have a specific CNTMT Ventilation isolation subsystem of ESFAS as some other plants do. This item meets the intent of the KA system in that it measures knowledge of those monitors that directly interface with the CNTMT purge system.
NUMBER PROCEDURE TITLE REVISION 31 1 -AP-5 UNIT 1 RADIATION MONITORING SYSTEM PAGE 3 of 5 -j-[ ACTiON! EXPECTED RESPONSE I I RESPONSE NOT OBTAINED
- 1. VERIFY THE FOLLOWING FOR THE AFFECTED RADIATION MONITOR(S): (Continued)
RADIATION MONITOR RECORDER ATT. NO. SGSTEAMLINE N/A ATTACHMENT 11 1-MS-RM-170, 171, 172 TURBINE DRIVEN AFW PUMP EXHAUST ATTACHMENT 11 N/A 1 -MS-RM-1 76 PERSONNEL HATCH AREA ATTACHMENT 6 1-RM-RR-100 1 -RM-RMS-1 61 MANIPULATOR CRANE 1-RM-RR-100 ATTACHMENT 5 1 -RM-RMS-1 62 CONTAINMENT ATTACHMENT 7 1-RM-RR-100 1 -RM-RMS-1 63 IN-CORE INST AREA 1 -RM-RR-1 00 ATTACHMENT 7 1 -RM-RMS-1 64 SG AND MAIN STEAM N-16 ATTACHMENT 2 1-MS-RR-193 1-MS-RI-190, 191, 192, 193 REACTOR COOLANT LETDOWN RADIATION 1-RM-RR-100 ATTACHMENT 8 MON RATEMETER, 1-CH-RI-128 DISCHARGE TUNNEL ATTACHMENT 4 1 -RM-RR-1 00 1 -SW-RM-1 30 SG BLOWDOWN 1-RM-RR-100 ATTACHMENT 9 1-SS-RM-124, 122, 123 CONTAINMENT PARTICULATE ATTACHMENT 5 1-RM-RR-100 1 -RM-RMS-1 59 (STEP 1 CONTINUED ON NEXT PAGE)
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-5 5 CONTAINMENT PARTICULATE, GASEOUS, AND REVISION MANIPULATOR CRANE RADIATION MONITORS PAGE 31 lof3 NOTE: 1-RMS-RM-162 is expected to be de-energized in modes 1,2, 3,4, and 5.
- 1. Do the following:
a) Have Health Physics determine if Containment gaseous and particulate samples are required. b) IF Containment gaseous and particulate samples are required, THEN have Health Physics obtain and analyze Containment gaseous and particulate samples. NOTE: If both containments are lined up for Containment Ventilation, then the Containment Vent Fans will restart when the Unit 1 Ventilation MOVs indicate full closed.
- 2. IF a Hi-Hi alarm is actuated OR a malfunction is suspected in Mode 5 or 6, THEN verify that Containment ventilation is isolated as indicated below:
- 1.-HV-F-4A-STOPPED D
- 1-HV-F-4B-STOPPED /
C
- 1-HV-F-5A-STOPPED c (2/O/4f C
- 1-HV-F-5B-STOPPED I4 C
- 1-HV-MOV-100A - CLOSED o C
- 1-HV-MOV-100B-CLOSED r (( /-te: v?7l--s C
- 1-HV-MOV-100C CLOSED-C
- 1-HV-MOV-100D CLOSED- I ii t
C
- 1-HV-MOV-1O1 -CLOSED C
- 1-HV-MOV-102 CLOSED
NUMBER ATTACHMENT TITLE ATTACHMENT 0-AP-5.2 3 1-GW-RI-178-1 2 OR 3 MGP PROCESS VENT RAD MONITOR REVISION PAGE 22 2of5
- 5. IF the abnormality of the Radiation Monitor was NOT caused by an obvious malfunction, THEN do the following for the MGP Process Vent Radiation Monitor(s) in alarm:
a) Inform the Health Physics Shift Supervisor of the following: Date and time the monitor alarmed and which channels are alarming
- Process Vent flow rate (1-GW-FI-108)
NOTE: The analysis should be completed immediately due to time limits for event classification in EPIP-1 .01, EMERGENCY MANAGER CONTROLLING PROCEDURE. b) Request a sample and survey of the affected area. c) IF a Hi alarm has actuated on 1 -G W-Rl-1 78-3 OR 1 -GW-RI-1 78-1, THEN verify the following: D
- 1-GW-TV-102A-CLOSED I
- 1-GW-TV-102B CLOSED G P -
D
- 1-GW-FCV-101 - CLOSED rv
- 1-CV-P-3A, Unit 1 Containment Vacuum Pump A STOPPED-C
- 1-CV-P-3B, Unit 1 Containment Vacuum Pump B - STOPPED C
- 2-CV-P-3A, Unit 2 Containment Vacuum Pump A STOPPED-C
- 2-CV-P-3B, Unit 2 Containment Vacuum Pump B STOPPED-d) Ensure the following trip valves are closed, as directed by the SRO, to stop potential radioactive sources to the Process Vent System:
C
- 1 -GW-TV-1 06, Equipment Vents C
- 1-GW-TV-1 13, Liquid Waste Tank Vents C
- 1-GW-TV-114, Boron Recovery Tank Vents e) Do the following to determine if the release has exceeded allowable limits:
- 1) Evaluate sample results, dose projections, and meter readings and compare to the limits in the EAL Matrix Categories as applicable for plant conditions, to determine the need to implement EPIPs.
- 2) IF EPIP implementation is NOT required, THEN initiate notifications specified in VPAP-2802, NOTIFICATIONS AND REPORT.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 31. 032AA2.06 03 1/BANKJ/L/3/3.9/4.1/7/
The Plant Computer System (PCS) is unavailable. In the event that the Main Control Room is evacuated, which ONE of the following identifies the instrument used to monitor reactor status, AND includes the location of the indicator? A. Excore Neutron Flux Monitor System Ch. I; Emergency Switchgear Room B. Excore Neutron Flux Monitor System Ch. I; Fuel Building C. Source Range Channel N-31 ; Emergency Switchgear Room D. Source Range Channel N-31 ; Fuel Building
- a. Incorrect. Correct instrumentation; location is incorrect but plausible, however the candidate who lacks detailed knowledge of App. R instrumentation would likely default to this location on the assumption that it would be desirable for the Unit RO to have direct indication at the Aux. S/D panel.
- b. Correct. First part correct as noted above; second part is also correct, the operator in the fuel building monitors and communicates with the controlling RO/SRO.
- c. Incorrect. First part is incorrect but plausible since this system is associated with accident monitoring and the candidate who does not have detailed knowledge, but knows that an accident isnt assumed for events involing control room evacuation, may assume that the normal instrumentation would be used. Second part incorrect but plausible as discussed in Distractor a.
- d. Incorrect. First part incorrect but plausible as discussed in Distractor c. Second part is correct as explained in answer b.
Loss of Source Range Nuclear Instrumentation Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation: (CFR 41.7 /45.5 /45.6) Confirmation of reactor trip Tier: 1 Group: 2 Technical
Reference:
0-FCA-1, TR 7.4 TBL 7.4-1, ICP-NM-1-NFD-190 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info: KA intent is met since knowledge is tested beyond that required by the KA. Additionally it was considered that if one asked soley about SR NIS being lost, using the excores (gamma metrics) would be fairly obvious, and it would be difficult to ask the question with 3 plausible distractors).
NUMBER PROCEDURE TITLE REVISION 36 O-FCA-1 CONTROL ROOM FIRE PAGE 18 of 70 STEP ACTION! EXPECTED RESPONSE J I RESPONSE NOT OBTAINED NOTE: The Fuel Building Auxiliary Monitoring Panel provides Appendix A indication.
- 18. SEND ONE LICENSED OPERATOR TO ESTABLISH REMOTE MONITORING USING ATTACHMENT 12, REMOTE MONITORING OPERATIONS
- 19. CONTACT AUXILIARY BUILDING 1J WHEN Steps of the applicable Attachment OPERATOR AND VERIFY THE FOLLOWING are complete, THEN GO TO Step 20.
ATTACHMENT STEPS COMPLETE: D Continue with Step 24. D
- ATTACHMENT 15, UNIT 1 REMOTE AUXILIARY BUILDING OPERATIONS, STEP 1 THROUGH STEP 4.i
- ATTACHMENT 14, UNIT 2 REMOTE AUXILIARY BUILDING OPERATIONS, STEP 1 THROUGH STEP 4.i
DZ m jnc UNIT 1 REMOTE MONITORING PARAMETERS C)
- 0) (I) 0 READING 0 PARAMETER 1 2 3 4 5 6* 7 8 LIMIT ACTION z
A SG LEVEL
>WRLeveI
(% WA Level) corresponding to Adjust Auxiliary 33% Narrow Feedwater Flow B SG LEVEL Range Level (% WA Level) NEUTRON FLUX 1 Stable or Align BIT m NEUTRON Decreasing FLUX 2 0
-1 m
RCSLOOPATc <555°F -1 Adjust Steam 0 H (° F) z Dumps or RHR 0 RCS LOOP B Tc As required by H I cooldown 0 cooldown curve (° F) m z z RCS LOOP ATh G) H 0 H (° F) Adjust Steam
<618°F -o H Dumps m I RCS LOOP B Th m
(° F) > H
< 2235 psig 0 PRESSURIZER Adjust z OR Cl)
PRESSURE Pressurizer As required by (psig) Heaters cooldown curve RCS Adjust RCS SUBCOOLING As Required Temperature or (° F) RCS Pressure PRESSURIZER Adjust Charging LEVEL 20%-29% and/or H (%) Seal injection H o > After column 6 is completed, notify Control room Operator to send blank copy of this log sheet to Fuel Building. m z H
NUCLEAR DESIGN INFORMATION PORTAL
- Appendix R Instrumentation 7.4 Table 7.4-1 (page 2 of 3)
Appendix R Instrumentation REQUIRED REQUIRED FUNCTION PRIMARY INSTRUMENT ALTERNATE INSTRUMENT LOOpS(a) LOOPS(a) Steam Generator Pressure 1-MS-PT-1474 to 1-MS-PI-111A, or 1-MS-PI1474, or 1-MS-PI-111B, or 1-MS-PT-1485 to 1-MS-PI-111C 1-MS-PI-1485, or 1-MS-PT-1496 to (2-MS-PT-2474 to (2-MS-PI-211A, or 2-MS-PI-2474, or 2-MS-PI-211B, or 2-MS-PT-2485 to 2-MS-PI-211C) 2-MS-PI-2485, or 2-MS-PT-2496 tq 2-MS-PI-2496) çb) RCS Pressure 1-RC-PT-1403 o 1-RC-PT-1000 to 1-RC-PI-14O3A 1-RC-PI-1000 (2-RC-PT-24O3 to (2-RC-PT-2000 to 2-RC-PI-24O3A and 2-RC-PI-2000) 2-RC-PT-2402 tQ 2-RC-PI-2402A) b) Pressurizer Level 1-RC-LT-1460 to 1-RC-LT-1000 to 1_RC_LI_1460(b) 1-RC-LI-1000 (2-RC-LT-2461 to (2-RC-LT-2000 to 2-RC-LI-2461) (b) 2-RC-LI-2000) Source Range Neutron Flux 1-NM-NFD-1270 t9 1-NM-NFD-19O t9 1-NM-NFI-127OA1, and 1-NM-NFI-19OA2 1-NI-NE-32 to 1-NI-NFI-32A and 32F (2-NM-NFD-2270 tç (2-NM-NFD-290 tQ 2-NM-NFI-2270A1) b) 2-NM-NFI-290A2) ib) (a) Numbers in parenthesis are Unit 2 instrument loops. (b) Instrumentation is RG 1.97 instrumentation as well as Appendix R instrumentation. Additional requirements of TRM Section TR 3.3.9 also apply. NAPS TRM 7.4-3 Rev 72, 03/18/09
PROCEDURE NO:
\\
\ VIRGINIAPOWER ICP-NM-1-NFD-190 REVISIONN UNITNO: NORTH ANNA POWER STATION 1 PROCEDURE TYPE: EFFECTIVE DATE: EXPIRATION DATE: INSTRUMENT CALIBRATION ON FILE N/A PROCEDURE TITLE: GAMMA-METRICS EXCORE NEUTRON FLUX MONITORING SYSTEM CHANNEL 1 (1-NM-NFD-190) CALIBRATION SURV APP RG REQ R EQ 1.97 REVISION
SUMMARY
NON-UPGRADED PROCEDURE
- For Isolator AT3 added data point of 10.300 based on shop feedback.
- For Isolator AT6 added data point of 10.300 based on shop feedback.
- Changed Level of Use from Continuous to Reference based on shop request.
ELECTRONIC DISTRIBUTION APPROVAL ON FILE REFERENCE USE
ICP-NM-l-NFD-190 Rev. 13 Page 3 of 16 2.0 Initial Conditions 2.1 Initial Plant Conditions 2.1.1 IF this procedure performance was driven by a Work Order for noise or spiking issues in Mode 5 or 6, THEN request the System Engineer to evaluate the channel prior to performance of this procedure or the Work Order. (Reference 13) 2.1.2 The Unit Supervisor has granted administrative authority for maintenance, testing or calibration to be performed and is aware of the following Alarms, Indications, Printouts, Recordings and Functions that will be affected: Indications: Indicator Description l-NM-NFI-l9OAl Source Range Count Rate (Reactor Parameter Indicator Assembly - Control Room) l-NM-NFI-190A2 Source Range Count Rate (Remote Monitoring Panel - Fuel Building) 1-NM--NFI-l9OBl Wide Range Power Level (Reactor Parameter Indicator Assembly - Control Room) l-NM-NFI-l90B2 Wide Range Power Level (Remote Monitoring Panel - Fuel Building) PCS Computer Point M1NMO13A EXCORE NEUTRON FLUX SR (RD) <> M1NMO14A EXCORE NEUTRON FLUX WR (RD) <> Alarms: Ann. 1A E-4 EXCORE NEUTRON FLUX MONITOR TROUBLE Functions: None
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 32. 033AA1 .01 032/BANKJNAPS/H/4/2.9/3.1/7/
Unit 1 is in Mode 2. A CONTROL POWER fuse blows on Intermediate Range Channel N-35. Which ONE of the following identifies the plant response for this event? A. N-35 detector will indicate downscale and the reactor will trip. B. N-35 detector will indicate normally and the reactor will trip. C. N-35 detector will indicate downscale and the reactor will NOT trip. D. N-35 detector will indicate normally and the reactor will NOT trip.
- a. Incorrect but plausible since the candidate who lacks detailed system knowledge may not be sure of the source of power to the detector (instrument or control power) and further may not recognize the change in bistable status.
- b. Correct. Detector high voltage is independent of control power (source is instrument power), so the failure will not effect indication. In this configuration the IR trip is not yet blocked (power is below P-i 0, < 10%) so a loss of control power will result in drawer bistable driver tripping. Thus a loss of control power will result in a reactor trip from 1/2 IR hi-flux.
- c. Plausible as discussed above. Even though the candidate may recall a difference between control &
instrument power fuse function they could easily confuse their operation, in which case they would choose this distractor.
- d. As noted above the candidate who lacks detailed system knowledge may confuse the function of the different drawer power supplies and failure modes and effects.
Loss of Intermediate Range Nuclear Instrumentation Ability to operate and I or monitor the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation: (CFR 41.7 / 45.5 / 45.6) Power-available indicators in cabinets or equipment drawers. Tier: 1 Group: 2 Technical
Reference:
ARs D-C3, A-A4, DWG 5655D33 sh 3 of 16 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
VIRGINIA POWER 1-EI-CB-21D ANNUNCIATOR C3 1-AR-D-C3 NORTH ANNA POWER STATION REV. 2 SNSOC APPROVAL: ON FILE Effective Date:05/06/02 NIS INTER Current equivalent to RGE HI FLUX 35% power on RX TRIP 1/2 Channels Interlocked with P10 1.0 Probable Cause 1.1 Uncontrolled RCCA bank withdrawal from a subcritical position 1.2 Failure to block permissive interlock Pb 1.3 Intermediate Range NI failure (c Cc 2.0 Operator Action 2.1 IF the Reactor is tripped, THEN GO TO 1EO, Reactor Trip or Safety Inj ection. 2.2 IF the Reactor is NOT tripped, THEN verify Reactor power level is normal by using alternate and redundant indications. 2.3 IF Reactor power level is NOT normal OR unable to determine that the Reactor is in a safe operating condition, THEN trip the Reactor and GO TO 1-E-0, Reactor Trip or Safety Injection. 2.4 IF Reactor power level is < 10% AND at least 1/2 NIS IR RX TRIP CHNL I / II (Panel LB1, B2) status alarms are LIT, THEN begin an orderly shutdown and initiate Actions in accordance with Tech Spec 3.0.3 (ITS 3.0.3) for inoperable Solid State Protection System. 2.5 IF Reactor power level is > 10% but < 35% AND the logic of Step 4.1 exists, THEN begin an orderly shutdown and initiate Actions in accordance with Tech Spec 3.0.3 (ITS 3.0.3) for inoperable Solid State Protection System. 2.6 IF Reactor power level is > 35% AND P10 PERM PR > 10% BLK NIS LP TRIPS (Panel PD2) is NOT LIT, THEN begin an orderly shutdown and initiate Actions in accordance with Tech Spec 3.0.3 (ITS 3.0.3) for inoperable Solid State Protection System. 2.7 IF all plant parameters are normal AND the alarm is due to unknown instrument malfunctions, THEN immediately contact Instrument Department to determine cause of alarm. 3.0 References
3.1 Precautions, Limitations, and Setpoints Document 3.2 Westinghouse Logic NADW5655D33 3.3 11715ESK1OD, 1OAAA 3.4 Tech Spec 3.0.3 (ITS 3.0.3) 3.5 1-E0, Reactor Trip or Safety Injection 3.6 DCP 97803, Intermediate Range NI Flux Trip Setpoint Change 4.0 Actuation 4.1 Reactor shutdown logic: EITHER of the following logic conditions satisfied:
- Pb PERM PR > 10% BLK NIS LP TRIPS (Panel PD2) is NOT LIT at least 1/2 NIS IR RX TRIP CHNL I / II (Panel LB1, B2) status alarms is LIT 4.2 1NINC35F 4.3 1NINC36F
-END-
COMPENSATED - IO CNAER DETECTOR ( I LOG AMPLIFIER NM2O1 AMPERES NEUTRON LEVEL II To IOAMPERES 3 MI 201 P0 OPERATION II8VAC HIGH SELECTOR SWITCH S201 I iLEVEE LEVEL RO STOP TO ROD STOP _..4 1b REMOTE PLANT BISTABLE II8VAC EQUIPMENT RELADRIVE RJ4 CONTROL R)wER (RPE) CALIBRATE IGH LEVEL ROD STOP MODULE NM203 ITRIP 118 VAC, HIGH LEVEL HIGH LEVa REACTOR TRIP 8ISTABLE TT MODE 1REYDRIVEJ..... 6 NTROLPOR (PROTECTION TO RPE CIRCUIT) 5202 H IGH LEVEL TRIP DS2O6 Pa POWER ABOVE LEVEL TRIP ABOVE SWITCH PERMISSIVE P6 520 TO RPE NC205 POWER ABOVE (PROTECTION
+ 25 V DC TO NM2OI, NC 201 PERMISSIVE PS CIRCUIT)
THROUGH NC2O OS?05 NN202 TRIP SPARE TRIP SPARE
+2SVOC BISTABLE TO RPE I\
SPARE 05VDC 0S204 D5VDC TO 0-sOwVoC 25V 25VDC TION AMPLIFIER OIOVDC TO 0IMA J NM2O? COMPARATOR AND RATE 0 TO 100 VDC ASSEMBLY U N31/N46 LOSS OF COMPENSATION II8VVCLOSS 01 COM PE N SAT I VOLTAGE COMPENSATION VOLTAGE VOlTAGE TO RPE SUPPL N 202 S Dl COMP
+311) TO +1500 VDC VOLT 05201 LOSS OF DETECTOR lB VC,LOSS OF
]. VOLTAGE DETECTOR VOLTAGE 118V, 50/60HZ 1 rTAGE 81 STABLE INSTRUMENT __41.j POWER SUPPLY L2O _______. TO RPE POWER L_2!__J RE S Of DETECTOR INSTRUMENT POWER ON P a 118V,50/6OHZ CONTROL 0S201 (SO? j LEVEL TRIP 118 VAC, LEVEL
- TRIP BYPASS II8V SWITCH POWER TO RPE 11 CONTROL. PdWEl (0 TRANSFORMERS 5203 (PROTECTION CONTROL OPERATION CIRCUIT)
CHANNEL TZOI THROUGH POWER ON SELECTOR T2O6 [FALL TRIP BYPASS DS2O8 ON TEST SWITCH S2OI 0S209 __ DS2IO CHANNEL ON TEST INDICATION TO RPE oH 130C057-.A Figure 25. Intermediate Range Circuit, Block Diagram 47c, 231 c- /i /I C) (j.
- 14) k /c
VIRGINIA POWER 1-EI-CB-21A ANNUNCIATOR A4 1 -AR-A-A4 NORTH ANNA POWER STATION REV. 1 APPROVAL: ON FILE Effective Date:6/20/O1 NIS IR LOSS OF DET 100V below nominal VOLTAGE 1.0 Probable Cause lAfL/ 0-f 1.1 Loss of high voltage to either or both intermddiate range instruments N-35 or N-36. 1c of Vt 2fYV 2.0 Operator Action 2.1 Refer to alarm lights on the front of the intermediate range drawers to determine affected instrument(s) 2.2 IF loss of detector voltage is confirmed, THEN GO TO 1AP--4.2, Malfunction of Nuclear Instrumentation (Intermediate Range) 3.0 References 3.1 UFSAR chapter 7 3.2 1AP4.2, Malfunction Of Nuclear Instrumentation (Intermediate Range) 3.3 117 15ESK1OAAK 3.4 Westinghouse drawing 271C351 sheet 27 3.5 Westinghouse NIS Tech Manual 3.6 Tech Spec 3.3.1.1 (ITS 3.3.1) Actuation 4.1 Bistable in intermediate range drawer N35 or N36 (NC 202)
INTEIMEDIATE RANGE REACTOR TRIP POWER RANGE REACTOR TRIP II II III IV POWER RANGE HIGH NEUTRON I II FLUX RATE REACTOR TRIP ( SOURCE RANGE BLOCK CONTROL NOTES 1 2 INTERMEDIATE RANGE BLOCK CONTROL (NOTES 1 g 3) I (NOTE 5) III ( NC 42 ()
-j-A T TRIP 2 BYPASS I/N 35A C ) (NIS RACK) 2/
C R HIGH NEUTRON FLUX MANUAL TRIP (HIGH SETPOINT) (MAIN CONTROL BOARD) REACTOR TRIP (SHEET 2) TO I.R. TO LR. RESET TRIP TRIP ROD STOP ROD STOP NOHEIIrARY (SHEET 4) (SHEET 4) H ONE NT ART TO I.R. HIGH NEUTRON FLUX BY I BY BY ROD STOP (SHEET 4) (LOW SETPOINT) REACTOR TRIP HEPS T jO OTHERS (SHEET 2) HIGH NEUTRON FLUX RESET REACTOR TRIP REACTOR TRIP REACTOR TRIP (SHEET 2) (SHEET 2) NOTES
- 1. THE REDUNDANT MANUAL BLOCK CONTROLS CONSIST OF TWO CONTROLS ON THE CONTROL BOARD FOR EACH RANGE, ONE FOR EACH TRAIN.
NUCLEAR STEAM SUPPLY SYSTEM
- 2. I/N 33A IS IN LOGIC TRAIN A. FUNCTIONAL DIAGRAMS I/N 330 IS IN LOGIC TRAIN B. NUCLEAR INSTRUMENT MANUAL TRIP SIGNALS
- 3. I/N 3BA IS IN LOGIC TRAIN A. UNITS 1 & 2 I/N 30B IS IN LOGIC TRAIN B.
- 4. I/N 47A IS IN LOGIC TRAIN A.
I/N 47B IS IN LOGIC TRAIN B.
- 5. TWO COMPUTER INPUTS ARE CONNECTED TO THIS CIRCUIT, INDIVIDUAL FOR EACH TRAIN.
- 6. MANUAL RESET CONTROLS CONSIST OF FOUR MOMENTARY CONTROLS IN THE CONTROL ROOM, ONE CONTROL FOR EACH INSTRUMENT CHANNEL.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 33. 034A4.0 1 033/BANK//H/4/3 .3/3.7/8/
Given the following conditions:
- Unit 1 has been shutdown for 6 days.
- Containment purge is in service on Unit 1.
- A containment air recirc fan (CARF) is running.
- Core off-load is in progress.
The A CARF trips, and the crew is unable to restart any CARE. Which ONE of the following describes impacts of this failure on the Radiation Monitoring System, AND on the core off-load in accordance with 1-OP-4.1, Controlling Procedure for Refueling? A. 1-RM-RMS-159 and -160, Containment Particulate and Gaseous monitors, are BOTH inoperable; Station Management approval is required in order to continue core off-load. B. 1-RM-RMS-160, Containment Gaseous monitor ONLY is inoperable; Station Management approval is required in order to continue core off-load. C. 1-RM-RMS-159 and -160, Containment Particulate and Gaseous monitors, are BOTH inoperable; Nuclear Analysis and Fuels approval is required in order to continue core off-load. D. 1-RM-RMS-160, Containment Gaseous monitor ONLY is inoperable; Nuclear Analysis and Fuels approval is required in order to continue core off-load.
- a. Correct. At least one CAREs must be in operation to support operability of the subject monitors.
Approval level is correct for the impact on 2 RMs during fuel movement.
- b. Incorrect. First part is incorrect but plausible since the candidate who lacks detailed knowledge may not understand the RM dependency on the CARES. Second part is correct as discussed above.
- c. Incorrect. First part is correct as discussed in answer a. Second part is incorrect but plausible since NAF is involved in the activity, however they do not have the approval authority for the given conditions.
- d. Correct. First part is incorrect as discussed in Distractor b; second part is also incorrect but plausible as discussed in distractor c.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Fuel Handling Equipment System (FHES) Ability to manually operate and/or monitor in the control room: (CFR: 41.7 I 45.5 to 45.8) Radiation levels Tier: 2 Group: 2 Technical
Reference:
1-OP-4.1 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
PROCEDURE NO: I -OP-4.1 REVISION NO: NORTH ANNA POWER STATION 64 PROCEDURE TYPE: UNIT NO: OPERATING PROCEDURE I PROCEDURE TITLE: CONTROLLING PROCEDURE FOR REFUELING SURV ICCE REACT REQ III MGT REVISION
SUMMARY
- FrameMaker Template Rev. 030.
Incorporated OP lO0038; Refueling procedure:
- Changed Attachment 9 by implementing lessons learned. This will save time and dose allowing better coordination between departments applying operational experience from the previous U -1 outage.
Incorporated OP-10-0093, Per John Slattery, delete P&L for conditions must be maintained when fuel is being moved in containment: System Engineering providing technical justification using screening process.
- Deleted bullet from Step 4.40, WHEN fuel is being moved within the Containment, THEN the following conditions must be maintained: (Reference 2.4.24)
- RWST level greater than 25 percent.
Administrative:
- Step 2.3.79 VPAP-2805 superseded by OU-AA-200.
- Step 4.1.2 DNAP-0509 replaced by AD-AA-102.
- Changed Step 4.4 continuation to include Step instructions in the next page continuation.
- Clarified Annunciator mark number.
- Added IV to IV initial lines where needed.
- Changed CR submitted to CR entered where appropriate.
- Changed Step 4.5.2 doubles to increases by a factor of 1.43 and 0.5 to 0.7. Ref 2.3.76 and 2.4.72. This change was made in a previous revision in Step 5.1.161, but Step 4.5.2 was inadvertantly omitted.
PROBLEMS ENCOUNTERED: j NO YES Note: If YES, note problems in remarks. REMARKS: (Use back for additional remarks.) SRO: DATE: CONTINUOUS USE
DOMINION 1-OP-4.1 North Anna Power Station Revision 64 Page 65 of 169 Reactor Vessel Upper Internals Removal 5.1.84 Refer to Attachment 15, Health Physics General Rules, Step 4 for additional guidance. 5.1.85 Verify the following have been done, using l-PT-96. 1, Refueling Systems Operability And Checkout, within 100 hours before commencing Core Alterations:
- Manipulator Crane Hoist Checks, in accordance with TRM TSR 3.9.3.1 and TSR 3.9.3.2.
- Manipulator Crane Auxiliary Hoist Load Test, in accordance with TRMTSR3.9.3.3.
I NOTE: IF no Containment Air Recirc Fans are in service, THEN 1-RM-RMS-159 and l--S-160 are NOT operable. ( (L \ NOTE: The Manipulator Crane Radiation Monitor is enabled in preparation for (<2 1 refueling activities and normally disabled at power to avoid damaging the C detectors. (Reference 2.4.28) 5.1.86 Reset the following instrumentation in accordance with TRM 3.3.7, Table 3.3.7-1, Items Lb and 2.b, and clear the Action Statement Status Log for movement of irradiated fuel assemblies within Containment:
- Manipulator Crane Area Monitor 1 -RM-RMS- 162 (Reference 2.4.28)
I&c
- Containment Particulate Monitor 1 -RM-RMS- 159
- Containment Gaseous Monitor 1 -RM-RMS-- 160 I&c
DOMINION 1-OP-4.1 North Anna Power Station Revision 64 Page 87 of 169 NOTE: 1-PT-93, Reactor Vessel Water Level Determination, is required to be done within 2 hours before the start of and at least once every 24 hours during the movement of Fuel Assemblies or Control Rods in accordance with Tech Spec SR 3.9.7.1 and TRM TSR 3.9.6.1. NOTE: A cavity level of 289 feet 10 inches is approximately 55 percent cold cal PRZR level indicated on 0l-RC-LI-1462. 5.1.156 Before moving fuel and at least once every 7 days thereafter, calibrate the Load Indicator for the Manipulator Crane using 1 -PT-96. 1.1, Pre Core Load Manipulator Crane Checkout. Time / Date calibrated: 5.1 .157 Ensure HP has installed High Rad barriers and have personnel in place to monitor radiation levels during the start of fuel movement. (Reference 2.4.65) 5.1.158 Notify HP to begin surveys and coverage as necessary for fuel movement. 5.1.159 Notify Reactor Engineering and NA&F that fuel movement will be starting within the following 4 hours.
DOMINION 1-OP-4.1 North Anna Power Station Revision 64 Page 118 of 169 (Page 3 of 6) Attachment 2 Core Alterations Checklist
- c. IF any secured fan discharge damper is NOT closed, THEN declare 1-RM-RMS-159 and 1-RM-RMS-160 inoperable and enter the TRM Actions for TR 3.3.7 and TR 3.9.5.
NOTE: TRM TR 3.3.7 requires that a Condition Report be entered within 12 hours if radiation monitor l-RM-RMS-159, 1-RM-RMS-160, or l-RM-RMS-162 is inoperable during movement of irradiated fuel within containment. An increased level of Station Management authorization is required to perform fuel movement as the defense-in-depth is reduced if one or more radiation monitors become inoperable. 6.4 Record the status of the following radiation monitors: 1 -RM-RMS- 159, Containment Particulate Monitor Operable NOT Operable 1 -RM-RMS- 160, Containment Gaseous Monitor Operable NOT Operable 1-RM-RMS- 162, Manipulator Crane Area Monitor Operable NOT Operable V 6.5 IF any radiation monitor listed in Step 6.4 is NOT Operable, THEN obtain 2 VCY appropriate Station Management approval to perform fuel movement: r c IF one of the three radiation monitors is NOT Operable, THEN obtain OPSMGR
) pennission from the Operations Manager to perform fuel movement.
i\. A
,_J IF o of the three radiation monitors are NOT Operable, THEN obtain DIR-O&M permission from the Director-Operations and Maintenance to perform fuel movement.
- IF all three of the radiation monitors are NOT Operable, THEN obtain FSRC permission from FSRC to perform fuel movement.
6.6 IF any radiation monitor listed in Step 6.4 is NOT Operable, THEN enter in the Action Statement Log to suspend fuel movement and obtain appropriate higher Station Management approval prior to continuing fuel movement if any of the remaining operable radiation monitors become inoperable.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 34. 036AK3 .01 034/BANK//L/3/3 .1/3.7/8/
Unit 2 is in a refueling outage with core on-load in progress. Given the below instruments, which ONE of the following identifies ALL of the instruments that can cause AUTOMATIC actuation of the Containment Evacuation alarm?
- 1) Source Range Channel N-31
- 2) Manipulator Crane Radiation Monitor, 2-RM-RMS-162
- 3) Containment Gaseous Radiation Monitor, 2-RM-RMS-160 A. 1&3 B. 2&3 C. 1 only D. 2 only
- a. Incorrect. Plausible since both conditions would be reasons to evacuate containment, but only item 1 can automatically actuate the alarm.
- b. Incorrect. Plausible since both conditions would be reasons to evacuate containment, but only item 1 can automatically actuate the alarm.
- c. Correct. As discussed above there is only 1 input which will actuate the alarm.
- d. Correct. Plausible since as discussed above increases in this monitor could be cause for evacuation; moreover it is only required in Mode 6, and the candidate who lacks detailed knowledge may assume that one of the reasons it is required is for the evac alarm function.
Fuel Handling Incidents Knowledge of the reasons for the following responses as they apply to the Fuel Handling Incidents: (CFR 41.5,41.10 /45.6 /45.13) Different inputs that will cause a reactor building evacuation Tier: 1 Group: 2 Technical
Reference:
1 -ICP-N l-N-3 1 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
VIRGINIA POWER l-EI-CB-2lA ANNUNCIATOR B2 l-AR-A--B2 NORTH ANNA POWER STATION REV. 3 APPROVAL: ON FILE Effective Date:12/08/09 NIS SR Set each outage at HI FLUX 5 times base count AT SHUTDOWN 1.0 Probable Cause 1.1 Increasing count rate while in source range. 1.2 Failure of a source range channel. 1.3 Neutron count increasing during start-up. 4 VI 2.0 Operator Action 2.1 IF in Mode 5 or 6, THEN do the following: 2.1.1 Ensure that audible alarm is present in containment OR make Gaitronics announcement to evacuate containment. 2.1.2 IF moving irradiated or fresh fuel assemblies within Containment, THEN do the following:
- a. Remove the last fuel assembly loaded to a safe location as directed by the Refueling SRO. (Ref. 3.6 and 3.7)
- b. Suspend Core Alterations.
2.1.3 Ensure that Containment is evacuated. 2.2 IF instrument has failed, THEN GO TO l-AP-4.l, Malfunction Of Source Range Nuclear Instrumentation. 2.3 IF due to reactor startup, THEN defeat the alarm. 2.4 IF counts are increasing and cause is NOT known, THEN initiate borat ion. 2.5 IF counts are increasing and cause is NOT known, THEN perform a shutdown margin in accordance with the 1-PT-b series, Shutdown Margin Determination. 3.0 References 3.1 VPAPl401 3.2 l-AP-4.l, Malfunction Of Source Range Nuclear Instrumentation 3.3 Precautions, Limitations, Setpoints document 3.4 Westinghouse NIS Tech Manual 3.5 Tech Spec 3.3.1, 3.9.3 3.6 b-OP-4.l, Controlling Procedure for Refueling 3.7 ET N-09-0082, Neutron Countrate Monitoring During Core Alterations 4.0 Actuation 4.1 Source range channel N-31, N32, comparator NC1O3
DOMINION 1-ICP-NI-N-31 North Anna Power Station Revision 8 Page 56 of 65 Data Table 6.11-3 High Flux At Shutdown Bistable Desired Acceptable Range CPS Setpoint As Found As Left State CPS CPS CPS MIN MAX Trip to Reset to 6.11.4 IF Reactor Power is less than the P-6 setpoint of 10-10 amps, THEN do the following:
- a. IF required, THEN have the OATC announce that the Containment Evacuation alarm will be tested.
- b. Place the HIGH FLUX AT SHUTDOWN switch in NORMAL.
PC
- c. Verify that Annunciator 1A B-i, NIS SR HI AT SD BLOCKED, goes OFF. (Annunciator will stay ON if N-32 is blocked.)
6.11.5 Place OPERATION SELECTOR switch in LEVEL ADJ. PC 6.11.6 While monitoring the HIGH FLUX AT SHUTDOWN front panel light, do the following:
- a. Adjust LEVEL ADJ potentiometer as necessary to obtain As Found Trip and Reset data and record in Data Table 6.11-3.
NOTE: Containment Evacuation alarm may be heard on Channel 6 of the Reactor Containment microphones.
- b. IF Reactor Power is less than the P-6 setpoint of 10- 10 amps, THEN verify the following operate properly:
- Containment Evacuation alarm (fl in Mode 2, 3, or 4 and alarm cannot be heard because of process noise, THEN mark this step N/A.)
- Annunciator 1A B-2, NIS SR HI FLUX AT SHUTDOWN
DOMINION 1-ICP-NI-N-31 North Anna Power Station Revision 8 Page 57 of 65 6.11.7 While monitoring the LEVEL TRIP front panel light, do the following:
- a. Adjust LEVEL ADJ potentiometer as necessary to obtain As Found Trip and Reset data, and record in Data Table 6.11-4.
- b. Verify the following operate properly:
- 1L A-i, NIS SR RX TRIP C1{NL I
- NOO3OD, SOURCE RANGE N31 HI FLUX Data Table 6.11-4 Level Trip Bistable Desired Trip Point: io CPS Allowable: 1.3 x iO CPS Desired Reset Point: 50% of setpoint Allowable: N/A Desired Acceptable Range CPS Setpoint As Found As Left State CPS CPS CPS MIN MAX Trip 10 4 7.7x10 to 5 l.3x10 Reset 4 5x10 4 3.9x10 to 4 6.5x10 6.11.8 IF any As Found data is NOT within the Acceptable Range or adjustments are desired, THEN do the following:
NOTE: Continuously bringing in the Containment Evacuation alarm while adjusting Bistables is disruptive to Containment Activities.
- a. Place the HIGH FLUX AT SHUTDOWN switch in BLOCK.
PC
- b. Make adjustments as needed.
- c. Place the HIGH FLUX AT SHUTDOWN switch in NORMAL.
PC
DOMINION 1-ICP-NI-N-31 North Anna Power Station Revision 8 Page 58 of 65
- d. While monitoring the HIGH FLUX AT SHUTDOWN and/or LEVEL TRIP front panel lights, do the following:
- Adjust LEVEL ADJ potentiometer as necessary to obtain As Left Trip and Reset data.
- Record the As Left data in the following data table(s):
Data Table 6.11-3 Data Table 6.11-4
- e. Ensure the following:
- For the HIGH FLUX AT SHUTDOWN bistable, IF Reactor Power is less than the P-6 setpoint of 10-10 amps, the following operate properly:
Annunciator 1A B-2, NIS SR HI FLUX AT SHUTDOWN Containment Evacuation alarm (IF in Mode 2, 3, or 4 and alarm cannot be heard because of process noise, THEN mark this step N/A.)
- For the LEVEL TRIP bistable, the following operate properly:
1LA-l,NIS SRRXTRIPCHNLI NOO3OD, SOURCE RANGE N31 HI FLUX 6.11.9 IF a new HIGH FLUX AT SHUTDOWN setpoint was installed, THEN ensure the date and time are recorded in the Reactor Data Book. 6.11.10 Rotate LEVEL ADJ potentiometer Ru 1 fully counterclockwise. 6.11 11
. Place the OPERATION SELECTOR switch in NORMAL.
PC 6.11.12 IF Reactor Power is less than the P-6 setpoint of iO° amps, THEN place PC the HIGH FLUX AT SHUTDOWN switch in BLOCK.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 35. 037AG2.1 .20 035/NEW//H14/4.6/4.6//
Given the following conditions:
- Unit 1 was at 100% power when a 50 gpm tube leak was identified in B SG.
- The crew is performing 1-AP-24, Steam Generator Tube Leak.
The Unit Supervisor is at Step 18, Verify Flow from Affected SGs - ISOLATED. The BOP reports that Attachment 3, MSVH Steam Generator Isolation Local Actions, has been initiated, but has NOT been completed. Which ONE of the following identifies the appropriate response, AND includes the reason? A. Continue with Step 19 and initiate RCS cooldown without delay; precludes overfill of B SG. B. Continue with Step 19 and initiate RCS cooldown without delay; minimizes offsite dose. C. Do NOT continue with Step 19, RCS cooldown can NOT be initiated until Attachment 3 is complete; ensures adequate delta-P is maintained between B SG and the unaffected SGs. D. Do NOT continue with Step 19, RCS cooldown can not be initiated until Attachment 3 is complete; ensures B SG PORV remains available.
- a. Incorrect. Plausible since potential SG overfill is a concern, but the need to have the SG isolated takes precedence since it is necessary in order to stop primary to secondary leakage.
- b. Incorrect. Plausible since offsite doses are a concern, but the need to have the SG isolated takes precedence since it is necessary in order to stop primary to secondary leakage..
- c. Correct. As already discussed the affected SG must be isolated in order to ensure pressure is maintained high enough above unaffected SGs to allow the RCS to be depressurized and equalized with the leaking SG while maintaining RCS subcooling.
- d. Incorrect. Action is correct as discussed in Distractor c. Basis is plausible since this condition is desired to avoid challenging a code safety, however the attachment only ensures that unaffected SG PORVs are available to ensure a steam release path for cooldown in the event that condenser steam dumps are unavailable or subsequently lost.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Steam Generator (S/G) Tube Leak Ability to interpret and execute procedure steps. (CFR: 41.10 /43.5/45.12) Tier: Group: Technical
Reference:
1-AP-24 and WOG background document Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
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STEP DESCRIPTION TABLE FOR ARG-3 STEP 15 STEP: Isolate Flow From Affected SG: PURPOSE: To isolate flow from the affected SG to minimize radiological releases and to maintain pressure in the affected SG greater than the pressure in the intact SGs following cooldown of the RCS in subsequent steps. BASIS: Isolation of the affected SG effectively minimizes release of radioactivity from this generator. In addition, isolation is necessary to establish a pressure differential between the affected and intact SGs in order to cool the RCS and stop primary-to-secondary leakage. This can be demonstrated by considering steady state energy transfer from the RCS to the SG5 simply expressed as: QRCS = 1 UA (TRCS TSGI) + UAA (TRCS TSGA) where QRCS = Heat generation rate in the primary system UA = Total convective heat transfer coefficient TRCS = Average RCS temperature SG1 = Intact SG SGA = Affected SG SUB I = refers to intact steam generator SUB A = refers to affected steam generator Assuming that the steam generators are at saturation conditions: QRCS = 1 [TRCS UA TSAT (PSGI)j + UAA [TRCS TSAT (PsGA)] where TSAT (P) = Saturation temperature at pressure, P The amount of subcooling in the primary system is expressed as b 5 T = TSAT (PRCS) - THOT In order to stop primary-to-secondary leakage, the primary pressure must be reduced to a value equal to that of the affected steam generator. In that case: PRCS = PSGA Tsb = TSAT (PSGA) - (TRCS + delta T/2) where delta T = core temperature rise. ARG-3 Background 113 HP/LP-Rev. 2, 4/30/2005 ARG3BD.doc
STEP DESCRIPTION TABLE FOR ARG-3 STEP 18 CAUTION CAUTION: Isolation of the affected SG from the intact SGs should be completed by closing the main steamline isolation and bypass valves for the affected 56 or for the intact SGs and closing the steam supply valve to the turbine-driven AFW pump from the affected SG before continuing to Step 18.Major steam flow paths from the affected SG should be isolated before initiating RCS cooldown. PURPOSE: To alert the operator that the affected steam generator should be isolated from the intact steam generators before steam is released from the intact steam generators in subsequent steps to cool the RCS. BASIS: Isolation of lall of the steam lines of the affected steam generator from the intact steam ginerators and isolation of the affected steam generator steam, supply valve to the turbine-driven AFW pump should be comred , decrr Lct nerr, presr;. V I AFW steam - Although cComplete isolétion of all large md small lines from, the affeded steam generator must be f,erformed expeditioUsly, butisolation of the small lines is not required prior to depressurization f the intr ill 11 i,will,, I ec)- cc KNOWLEDGE: Isolation of steam flow from the affected steam generator through branch lines off the main steamlines should be completed expeditiously, but is not required prior to RCS cooldown. Isolation of the branch lines may be completed in parallel with the RCS cooldown and subsequent recovery steps. PLANT-SPECIFIC INFORMATION: ARG-3 Background 121 HP/LP-Rev. 2, 4/30/2005 ARG3BD. doc
NUMBER PROCEDURE TITLE REVISION 20 1 -AP-24 STEAM GENERATOR TUBE LEAK PAGE 5of23 STEP ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED
- 3. IDENTIFY AFFECTED SG:
(Continued) D
- Steam Flow-Feed Flow mismatch OR D
- MFRV position OR D
- Local Main Steamline contact radiation readings 4._ ISOLATE FLOW FROM AFFECTED SG:
D a) Place affected SG PORV Controller \-/ setpoint at 1050 psig (pot setting at 5.6) b) Check affected SO PORV - D b) WHEN affected SG pressure is less than CLOSED: 1050 psig, THEN verify SG PORV is closed. D
- 1-MS-PCV-1O1A D IF PORV is NOT closed, THEN manually close P0 RV.
D
- 1-MS-PCV-1O1B IF PORV cannot be closed, THEN locally close D
- 1-MS-PCV-1O1C PORV:
D
- 1-MS-PCV-1O1A D
- 1-MS-PCV-1O1B D
- 1-MS-PCV-1O1C (STEP 4 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 20 1 -AP-24 STEAM GENERATOR TUBE LEAK PAGE 6 of 23 -f_STEP ACTION/EXPECTEDRESPONSE I RESPONSENOTOBTAINED I
- 4. ISOLATE FLOW FROM AFFECTED SG: (Continued)
C c) Close affected SG MSTV and Bypass Valve C d) Initiate ATTACHMENT 3, MSVH STEAM GENERATOR ISOLATION LOCAL ACTIONS C e) Verify affected SG Blowdown Trip e) IF affected SG Blowdown Trip Valves are NOT Valves CLOSED open for sampling procedure, THEN manually close affected SG Blowdown Trip Valves: C
- 1-BD-TV-100A (A SG)
C
- 1-BD-TV-100B (A SG)
C
- 1-BD-TV-100C (B SG)
C
- 1-BD-TV-100D (B SG)
C
- 1-BD-TV-100E (C SG)
C
- 1-BD-TV-100F (C SG)
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-24 3 MSVH STEAM GENERATOR ISOLATION LOCAL ACTIONS REVISION PAGE 20 lof4 STEP ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED
- 1. CLOSE AFFECTED SOs STEAM SUPPLY VALVE TO THE TURBINE-DRIVEN AFW PUMP L (TERRY-TURBINE):
C
- 1-MS-18, A Main Steam V Header To Turb Drvn AFW Pp Isol Vv C
- 1 -MS-57, B Main Steam 2 Header To Turb Drvn AFW Pp Isol Vv C
- 1-MS-95, C Main Steam Header To Turb Drvn AFW Pp lsol Vv
- 2. VERIFY ALL NON-AFFECTED SG C HZ at least one NON-AFFECTED SO PORV PORV ISOLATION VALVES OPEN:
Isolation Valve is open, THEN GO TO Step 3. C
- 1-MS-21, 1-MS-PCV-1O1A HZ at least one NON-AFFECTED SO PORV Isolation Valve Isolation Valve is NOT open, THEN do the following:
C
- 1-MS-59, 1-MS-PCV-1O1B Inlet Isolation Valve C a) Open one NON-AFFECTED SG PORV Isolation.
C
- 1-MS-97,1-MS-PCV-lOlCInlet Isolation Valve C b) GOTO Step 3.
3._ NOTIFY CONTROL ROOM THAT AFFECTED SGs ARE ISOLATED AND NON-AFFECTED SG(s) ARE AVAILABLE FOR COOLDOWN - ATTACHMENT 3, STEP 3-COMPLETE
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-24 3 MSVH STEAM GENERATOR ISOLATION LOCAL ACTIONS REVISION PAGE 20 2of4 f_STEP_ff ACTION/EXPECTED RESPONSE I RESPONSE NOTOBTAINED I NOTE: At least two Non-affected SG PORVs should be available to support cooldown and depressurization. ,)7
- 4. CHECK ALL NON-AFFECTED SG D Open NON-AFFECTED SG PORV Isolation Valve.
PORV ISOLATION VALVES OPEN:- D
- 1-MS-21, 1-MS-PC V-lOlA WHEN NON-AFFECTED SG PORV Isolation Isolation Valve Valve is open, THEN GO TO Step 5.
D
- 1-MS-59, 1-MS-PCV-1O1B Inlet Isolation Valve D
- 1-MS-97, 1-MS-PCV-1O1C Inlet Isolation Valve
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-24 3 MSVH STEAM GENERATOR ISOLATION LOCAL ACTIONS REVISION PAGE 20 3of4 ACTION/EXPECTED RESPONSE RESPONSE NOTOBTAINED _STEP 5._ UNIT IN MODE 1 OR 2 AT THE D IF Unit was in Mode 3 at the start of the event, START OF THE EVENT THEN do the following to ensure the affected SG traps are isolated (located on 1St floor): a) IF A SG is affected, THEN ensure the following valves are closed: D
- 1-MS-27, 1-MS-TV-lOlA Drain Header Isolation Valve
- 1-MS-32, 1-MS-TD-5 Inlet Isolation Valve 1I
- 1-MS-23, A Main Steam Header Blowdown Valve b) IF B SG is affected, THEN ensure the following valves are closed:
D
- 1-MS-65, 1-MS-TV-biB Drain Header Isolation Valve D
- 1-MS-70, i-MS-TD-7 Inlet Isolation Valve
- 1 -MS-61, B Main Steam Header Blowdown Valve c) IF C SG is affected, THEN ensure the following valves, are closed:
- 1-MS-103, 1-MS-TV-i O1C Drain Header Isolation Valve 1J
- 1-MS-b08, 1-MS-TD-9 Inlet Isolation Valve
- 1-MS-99, C Main Steam Header Blowdown Valve
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QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 36. 038EA2.05 036/BANKJ/H/3/2.8/2.9/3/
Given the following conditions:
- Unit 1 was at 100% power when a SG tube rupture occurred on A SG.
- The crew is performing 1-E-3, Steam Generator Tube Rupture.
SG narrow-range levels are as follows:
- A SG is offscale high.
- B SG is 22%.
- C SG is 24%.
Which ONE of the following describes the method the crew will use to mitigate this event? A. Cooldown the RCS at< 100°F/hr, then depressurize the RCS using maximum available PRZR spray. B. Cooldown the RCS at< 100°F/hr, then depressurize the RCS using one PRZR PORV. C. Cooldown the RCS at the maximum controllable rate, then depressurize the RCS using maximum available PRZR spray. D. Cooldown the RCS at the maximum controllable rate, then depressurize the RCS using one PRZR PORV.
- a. Incorrect. Plausible since other procedures (e.g. 1-AP-24, 1-ES-1.2, etc.) stipulate the 100°F/hr rate and since the question does not raise issues over offsite doses and the like, therefore is no implied urgency. Second part is also incorrect but plausible since this is the preferred method of the EOP to minimize RCS inventory loss; for this case where ruptured SG level is offscale high the alternative method (Step 16c, RNO) is utilized, however the candidate who lacks detailed knowledge of the procedure and step basis may default to this distractor.
- b. Incorrect. First part incorrect but plausible as noted above. Second part is correct, although most likely there will be no challenge regarding SG overfill concerns, the procedure conservatively takes this approach in that regard with the goal being to terminate primary-to-secondary leakage in the most expeditious fashion.
- c. Incorrect. First part is correct; as alluded to previously rather than stipulating conditions (which would complicate the procedure), the EOP selects this method with the goal of terminating primary-to-secondary leakage in the shortest reasonable time frame. Second part incorrect but plausible as discussed in Distractor a.
- d. Correct. First part is correct as discussed in Distractor c. Second part is also correct as discussed in Distractor b.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Steam Generator Tube Rupture (SGTR) Ability to determine or interpret the following as they apply to a SGTR: (CFR 43.5 /45.13) Causes and consequences of shrink and swell in S/Gs Tier: 1 Group: 1 Technical
Reference:
EOP E-3, 1-AP-24, 1-AP-24.1 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info: Considered to meet K/A intent since knowledge level is beyond that asked by the K/A; Additionally, for the given conditions (SGTR in progress) the response of SG level to initiating and subsequently terminating steam dumping from intact SGs (swell and shrink respectively) seemed fairly obvious, thus making it difficult to construct a question at the RO level with 3 plausible distractors.
NUMBER PROCEDURE TITLE REVISION 26 1 -E-3 STEAM GENERATOR TUBE RUPTURE PAGE 11 of 45 H_STEP ACTION/EXPECTED RESPONSE RESPONSE NOTOBTAINED I CAUTION:
- If no RCPs are running, then cooldown and depressurization may give a false Cold Leg temperature indication on the ruptured loop. This indication should be disregarded for the Integrity Status Tree until after performance of ATTACHMENT 3, OPERATOR ACTIONS TO CONTROL RCS PRESSURE AND CHARGING FLOW TO MINIMIZE RCS-TO-SECONDARY LEAKAGE.
If Steam Dumps are used for RCS cooldown, then, to prevent an undesired Main Steamline Isolation, each Main Steamline flow should be kept less than 1 .0E6 LBM/HR.
- 12. INITIATE RCS COOLDOWN:
EJ a) Determine required core exit temperature based on SG pressure: LOWEST RUPTURED SG CORE EXIT PRESSURE (PSIG) TEMPERATURE (°F) 1100 51 5°F [465°F] 1000 AND < 1 100 500°F [450°F] 900 AND < 1000 490°F [440°F] 800 AND < 900 475°F [425°F] 700 AND < 800 460°F [41 0°F] 600 AND < 700 445°F [395°F] 500 AND < 600 425°F [375° F] 400 AND < 500 405°F [355°F] 350 AND < 400 390°F [340°F] (STEP 12 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 26 1 -E-3 STEAM GENERATOR TUBE RUPTURE PAGE 12 of 45 -_STEP ACTION/EXPECTEDRESPONSE I I RESPONSENOTOBTAINED I
- 12. INITIATE RCS COOLDOWN: (Continued)
D b) Verify Condenser Steam Dumps - b) Manually or locally dump steam from AVAILABLE: intact SGs at maximum controllable rate using:
- 1) Transfer Condenser Steam Dump to Steam Pressure Mode:
- SG PORVs OR D a. Put both Steam Dump Interlock switches to OFF/RESET
- Decay Heat Release Valve:
D b. Put Steam Dump Controller to a. Locally open isolation valve(s) for MANUAL NON-RUPTURED SG(s) to Decay Heat Release Valve: D c. Put Mode Selector switch to STEAM PRESS
- 1-MS-19, A Steam Line to 1-MS-HCV-104 Non-Return D d. Verify or reduce Steam Dump Valve demand to zero D
- 1 -MS-58, B Steam Line to D e. Put both Interlock switches to 1 -MS-HCV-1 04 Non-Return ON Valve D
- 1 -MS-96, C Steam Line to 1 -MS-HCV-i 04 Non-Return Valve
- b. Locally open 1 -MS-20, Decay Heat Release Valve Upstream Isolation Valve.
- c. Open i-MS-HCV-104, Decay Heat Release Valve.
(STEP 12 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 26 1 -E-3 STEAM GENERATOR TUBE RUPTURE PAGE 13 of 45 -_STF} ACTION! EXPECTED RESPONSE _f RESPONSE NOT OBTAINED
- 12. INITIATE RCS COOLDOWN: (Continued)
IF no intact SG is available, THEN do the following: D
- Use faulted SG OR
- GO TO 1 -ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT SUBCOOLED RECOVERY DESIRED, STEP 1.
D 2) Check Panel P-F3, P-12 PERM D 2) Raise Steam Dump Controller TAVE <543F BLOCK SI BYP STM demand and Dump steam to DUMPS LIT
- Condenser from intact SGs at maximum controllable rate.
D
- Hold both Steam Dump lntlk Switches in BYP INTK until Panel P-F3 is LIT, then continue with Step 12b3.
OR D
- WHEN Panel P-F3 is LIT, THEN continue with Step 12b3.
D 3) Verify Panel P-F4, STM DUMP D 3) Momentarily place both Steam Dump COOLDOWN VLV AVAILABLE LIT - Intlk Switches in BYP INTK. Verify Panel P-F4 - LIT. D 4) Raise Steam Dump Controller demand and Dump steam to Condenser from intact SGs at çi1 1 maximum controllable rate f-v (STEP 12 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 26 1 -E-3 STEAM GENERATOR TUBE RUPTURE PAGE 14 of 45 -_STEP ACTION/EXPECTED RESPONSE I RESPONSE NOTOBTAINED
- 12. INITIATE RCS 000LDOWN: (Continued)
D c) Verify Panel P-G3, STM LINE PRESS - D c) Momentarily place both High Steam Flow FLOW SI BLKD TRNS A&B LIT - SI Block Switches in BLOCK. Verify panel P-G3 - LIT. d) Core exit TCs LESS THAN REQUIRED
- d) Continue RCS cooldown.
TEMPERATURE Do NOT proceed with Step 12e until core exit TCs less than required temperature. D e) Stop RCS cooldown
- 1) Maintain core exit TCs LESS THAN REQUIRED TEMPERATURE CHECK RUPTURED SGs PRESSURE - H ruptured SG pressure is at least 250 psi STABLE OR INCREASING greater than pressure of highest intact SG used for cooldown, THEN do the following:
D a) Check ruptured SG isolation. D IF SG is NOT isolated, THEN complete actions in Step 3. b) Initiate cooldown at less than 100°F/hr using intact SGs to maintain ruptured SG pressure greater than 250 psi above pressure of highest intact SG used for cooldown. IF pressure continues to decrease to less than 250 psi above intact SGs pressure used for cooldown, THEN GO TO 1-ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT SUBCOOLED-RECOVERY DESIRED, STEP 1.
NUMBER PROCEDURE TITLE REVISION 26 1 -E-3 STEAM GENERATOR TUBE RUPTURE PAGE 15 of 45 STEP ACTION/EXPECTEDRESPONSE I I RESPONSENOTOBTAINED I CHECK RCS SUBCOOLING BASED ON 1J GO TO 1-ECA-31, SGTR WITH LOSS OF CORE EXIT TCs GREATER THAN
- REACTOR COOLANT SUBCOOLED 45°F [95°F] RECOVERY DESIRED, STEP 1.
15._ BLOCK LOW PRZR PRESS SI SIGNAL: D a) Check PRZR pressure - LESS THAN EJ a) WHEN PRZR pressure less than 1950 PSIG 1950 psig, THEN perform Step 15b. D GOTOStep16. D b) Verify Panel P-G4, PRZR LO-LO PRESS D b) Momentarily place both Low PRZR SI BLKD TRNS A & B LIT - Pressure SI Block Switches in BLOCK. D Verify Panel P-G4 - LIT. NOTE: ATTACHMENT 5, CONDITIONS FOR TERMINATION OF RCS DEPRESSURIZATION, provides a handout for the substep conditions of Step 1 6e and Step 1 7b.
- 16. DEPRESSURIZE ACS TO MINIMIZE BREAK FLOW AND REFILL PRZR:
a) Place PRZR Heaters in PTL D b) Ruptured SGs narrow range level - ON D b) IF a PRZR PORV is available, THEN SCALE GO TO Step 1
/Jb HaPRZRPORVisNOTavailable, THEN GOTO Step 16c.
c) Normal PRZR spray AVAILABLE
- 1J c) GOTO Step 17.
d) Spray PRZR with maximum available spray (STEP 16 CONTINUED ON NEXT PAGE)
[ NUMBER PROCEDURE TITLE REVlON 1 -E-3 STEAM GENERATOR TUBE RUPTURE PAGE 16 of 45
-j_STEP f] ACTION/EXPECTEDRESPONSE 1 RESPONSENOTOBTAINED I
- 16. DEPRESSURIZE RCS TO MINIMIZE BREAK FLOW AND REFILL PRZR:
(Continued) e) Verify PRZR pressure satisfactorily e) Close Normal PRZR Spray Valves using decreasing until any of the following controller: conditions satisfied: D
- 1-RC-PCV-1455A
- PRZRIeveI-GREATERTHAN
- 1-RC-PCV-1455B 69%[50%]
OR D Verify PRZR spray valves closed. IF NOT, THEN place failed valve remote close D
- RCS subcooling based on Core Exit switch in CLOSE:
TCs LESS THAN 25°F [75°F] EJ
- 1-RC-SOV-1455A, 1-RC-PCV-1455A REMOTE CLOSE Both of the following: SOV D 1) RCS pressure LESS THAN D
- 1RCSOV1455B, RUPTURED SGs PRESSURE 1RCPCV1455B REMOTE CLOSE SOV D 2) PRZR level - GREATER THAN 21% [26%] IF spray valves can NOT be closed, THEN do the following:
HZ 1-RC-PCV-1455A failed open, THEN: D 1) Stop 1-RC-P-1C. D 2) Stop 1-RC-P-1A. flZ 1-RC-PCV-1455B failed open, THEN: D 1) Stop 1-RC-P-1C. D 2) IF 1-RC-P-1A is running, THEN stop 1 -RC-P-1 B. D GOTOStep17. (STEP 16 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 26 1 -E-3 STEAM GENERATOR TUBE RUPTURE PAGE 17 of 45 STEP ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED
- 16. DEPRESSURIZE RCS TO MINIMIZE BREAK FLOW AND REFILL PRZR:
(Continued) f) Close Normal PRZR Spray Valves: f) Place failed valve remote close switch in CLOSE: D
- 1-RC-PCV-1455A D
- 1-RC-SOV-1455A, D
- 1-RC-PCV-1455B 1-RC-PCV-1455A REMOTE CLOSE SOV D
- 1-RC-SOV-1455B, 1-RC-PCV-1455B REMOTE CLOSE SOV IF spray valves can NOT be closed, THEN do the following:
HZ 1-RC-PCV-1455A failed open, THEN: D 1) Stop 1-RC-P-1C. Ei 2) Stop 1-RC-P-1A. HZ 1-RC-PCV-1455B failed open, THEN: D 1) Stop 1-RC-P-1C. D 2) IF 1-RC-P-1A is running, THEN stop 1-RC-P-1B. IJ g) GOTO Step 19
NUMBER PROCEDURE TITLE REVISION 26 1 -E-3 STEAM GENERATOR TUBE RUPTURE PAGE 18 of 45 J_STEP_jj ACTION/EXPECTED RESPONSE RESPONSE NOTOBTAINED I CAUTION:
- The PRT may rupture if a PRZR PORV is used to depressurize the RCS.
This may result in abnormal Containment conditions. To prevent possible valve failure, cycling of PRZR PORVs should be minimized. NOTE: The Upper Head region may void during RCS depressurization if RCPs are not running. This will result in a rapidly increasing PRZR level.
- 17. DEPRESSURIZE RCS USING PRZR PORV TO MINIMIZE BREAK FLOW AND TO REFILL PRZR:
D a) PRZR PORV AT LEAST ONE
- 1J a) H normal Spray is available, THEN AVAILABLE RETURN TO Step 16.d.
H NOT THEN, GOTO 1-ECA-3.3, SGTR WITHOUT PRESSURIZER PRESSURE CONTROL, STEP 1. (STEP 17 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 26 1 -E-3 STEAM GENERATOR TUBE RUPTURE PAGE 19 of 45 -_STEP ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED
- 17. DEPRESSURIZE RCS USING PRZR PORV TO MINIMIZE BREAK FLOW AND TO REFILL PRZR: (Continued) b) Open one PRZR PORV until any of the D b) IF normal Spray is available, THEN following conditions satisfied: RETURN TO Step 16.d.
- PRZR level - GREATER THAN D NOT THEN, GO TO 1-ECA-3.3, SGTR 69% [50%] WITHOUT PRESSURIZER PRESSURE CONTROL, STEP 1.
OR D
- RCS subcooling based on Core Exit TCs LESS THAN 25°F [75° F]
OR
- Both of the following:
D a. RCS pressure LESS THAN RUPTURED SGs PRESSURE D b. PRZRIevel-GREATERTHAN 21% [26%] C c) Close PRZR PORV C c) Close PRZR PORV Block Valve.
NUMBER PROCEDURE TITLE REVISION 20 1 -AP-24 STEAM GENERATOR TUBE LEAK PAGE 16 of 23 -_STEP_j ACTION/EXPECTED RESPONSE RESPONSE NOTOBTAINED I NOTE: Charging flow should be adjusted to compensate for PRZR level decrease due to RCS cooldown.
- 20. INITIATE RCS COOLDOWN:
D a) Determine required core exit C temperature based on SO pressure: LOWEST AFFECTED SO CORE EXIT PRESSURE (PSIG) TEMPERATURE (°F) 1100 515 1000AND<1100 500 900AND<1000 490 800AND<900 475 700AND<800 460 600AND<700 445 500AND<600 425 400AND<500 405 35OAND<400 390 (STEP 20 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 20 1 -AP-24 STEAM GENERATOR TUBE LEAK PAGE 17 of 23 -_STEP ACTION/EXPECTEDRESPONSE I I RESPONSENOTOBTAINED
- 20. INITIATE RCS COOLDOWN:
(Continued) D b) Dump steam to Main Condenser D b) Manually or locally dump steam at a controllable from intact SGs at a controllable rate less than or equal to 100°F/hr from intact rate of 100°F/hr or less SG PORVs. D c) Verify core exit TCs LESS THAN
- 1 c) RETURN TO Step 20b.
REQUIRED TEMPERATURE D d) Stop RCS cooldown and maintain required RCS temperature
- 21. DEPRESSURIZE RCSTO STOP BREAKFLOW:
D a) Check normal PRZR spray - a) Do the following: AVAILABLE D 1) IF letdown is NOT in service, THEN place letdown in service.
- 2) Establish Auxiliary Spray as follows:
E1 a. Open 1 -CH-HCV-1 311, Auxiliary Spray Valve.
- b. Close 1-CH-HCV-1310, Normal Charging Isolation Valve.
D 3) Monitor PRZR level. D b) Spray PRZR with maximum available spray (STEP 21 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 24 1 -AP-24 1 SHUTDOWN STEAM GENERATOR TUBE LEAK
. PAGE 10 of 16
-_STEP_FI ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED NOTE: Tech Specs 3.4.13.d, 3.4.20, and 5.5.8 are applicable for identified SG tube leakage in Mode 4. 15.. NOTIFY MANAGER OF NUCLEAR OPERATIONS OR OPERATIONS MANAGER ON CALL THAT COOLDOWN AND DEPRESSURIZATION WILL BE COMMENCING
- 16. INITIATE RCS COOLDOWN AT 75 °F/HOUR OR LESS USING ANY OF THE FOLLOWING: C, LJ
- RHR System D
- Steam Dump to Condenser D
- Intact SG PORVs
- 17. CHECK IF SI ACCUMULATORS SHOULD BE ISOLATED:
a) Check RCS pressure LESS - D a) WHEN pressure is less than 1000 psig, THEN THAN 1000 psig perform Step 17b. b) Isolate SI Accumulators: D b) Vent any SI Accumulator that cannot be isolated using 1-OP-7.3, FILLING, SLUICING,
- 1) In1t1ateATTACHMENT3to DRAINING, PRESSURIZING, AND VENTING SI locally restore power to SI ACCUM ULATORS.
Isolation Valves
- 2) Close all SI Accumulator Discharge Isolation Valves:
D
- 1-Sl-MOV-1865A D
- 1-Sl-MOV-1865B
- 1-SI-MOV-1865C
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 37. 039A4.O 1 037/BANK!/L/3/2.9/2.8/4/
In accordance with 1-OP-28.3, Startup of the Moisture Separator Reheaters, when placing the Reheat Steam System in service, the total rate of change of the LP Turbine Inlet (Reheater Outlet) temperature should NOT exceed This heatup rate is initially controlled by A. 100°F per hour; throttling the MSR FCVs from the Benchboard B. 100°F per hour; throttling the 3-inch bypass valves locally C. 150°F per hour; throttling the MSR FCVs from the Benchboard D. 150°F per hour; throttling the 3-inch bypass valves locally
- a. Incorrect. Rate is correct per the OP; method is incorrect but plausible since the candidate who lacks integrated knowledge would most likely choose using the benchboard, since this is the way they are removed from service during a normal shutdown or fast load reduction.
- b. Correct. As mentioned the rate is correct and also as stated in the procedure Precaution and Limitation 4.2; as discussed above this is the prescribed method for the plant conditions (startup) and done primarily for finer control.
- c. Incorrect. First part is incorrect but plausible if the candidate erroneously associates the 50°F step change limit as non-inclusive with the normal 25°F per 15/mm (50+ 25 X 4 = 150). Second part incorrect but plausible as discussed in Distractor a.
- d. Incorrect. First part is incorrect but plausible as discussed in Distractor c. Second part is correct as discussed in answer b.
Main and Reheat Steam System (MRSS) Ability to manually operate and/or monitor in the control room: (CFR: 41.7 /45.5 to 45.8) Main steam supply. valves Tier: 2 Group: 1 Technical
Reference:
1-OP-28.3 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
DOMINION 1-OP-28.3 North Anna Power Station Revision 16 Page 4 of 27 2.4 Commitment Documents 2.4.1 CTS Assignment 02-91-0208, close the 1 inch warmup isolations for the MSRs during preparation for startup. 2.4.2 Safety Evaluation 99-SE-TM-08, Reheat Control Panel mit Verif 3.0 INITIAL CONDITIONS Review the equipment status to verify station configuration supports the performance of this procedure. 4.0 PRECAUTIONS AND LIMITATIONS 4.1 Comply with the following guidelines when marking steps N/A:
- IF the conditional requirements of a step do not require the action to be performed, THEN mark the step N/A.
- IF any other step is marked N/A, THEN have the SRO (or designee) approve the N/A and justify the N/A on the Procedure Cover Sheet.
DOMINION 1-OP-28.3 North Anna Power Station Revision 16 Page 5 of 27 4.2 WHEN placing the Reheat Steam System in service, THEN observe the following: (Reference 2.3.4)
- Due to thermal stresses and possible distortion of LP Turbine stationary parts, the maximum instantaneous temperature change of steam to the LP Turbine inlet (Reheater outlet temperatures) should be limited to 50°F. Including such an instantaneous change, the total rate of change of this temperature should NQI exceed 100°F per hour.
- Reheater Control System LP 1 and LP 2 are the more accurate indication of LP Turbine inlet temperatures, NOT individual Reheater outlet temperatures.
())
- IF the LP Turbine inlet temperature limitations are exceeded, THEN a Condition Report should be entered AND the System Engineer should be contacted for Turbine cyclic life evaluation.
4.3 Reheater outlet steam temperatures (LP Turbine inlet temperatures) MUST be closely monitored The heat-up rate OR cool-down rate limit of 25°F in a 15-minute period MUST NOT be exceeded. 4.4 To avoid the damage caused by water hammer when placing the Reheater Warm-up System in service, the throttle valves MUST be throttled open in very SMALL increments AND sufficient time MUST be allowed between valve movements to verify temperature changes. 4.5 To prevent potential Turbine damage, the Unit should NOT be operated at low loads for extended periods without the MSRs in service. 4.6 IF Unit 1 Reactor trips during MSR warmup, THEN the throttle valves on the 3 inch warmup lines MUST be isolated in accordance with Attachment 1, Actions for Reactor Trip During MSR Warmup, to prevent unwanted cooldown. 4.7 Before placing the MSRs in service, the load on the Main Turbine is less than or equal to 35 percent of rated capacity. IF load greater than 35 percent, THEN consult Westinghouse for instructions.
DOMINION 1-OP-28.3 North Anna Power Station Revision 16 Page 6 of 27 4.8 WHEN changing the alignment of the Moisture Separator Reheater (MSR) Drain Tank vent valves, THEN observe the following precautions:
- To avoid the damage caused by water hammer when opening the MSR Drain Tank vent valves, the valves MUST be throttled open in very SMALL increments AND sufficient time MUST be allowed between valve movements.
- The potential for a water hammer event increases with the length of time the vent valve is closed. If the vent valve has been closed less than one hour, the potential for a water hammer is minimized. If the vent valve is closed for one hour or more, the probability for a water hammer event is much more likely. In this case, extreme caution should be used when re-opening a vent valve.
- System Engineering should be notified prior to performing the procedure to determine if Instrument Department support will be required to instrument and collect data on the MSR drain pressures and MSR Drain Tank level.
49 Reactor power changes due to Steam Flow changes will affect Tavg
DOMINION 1 -OP-28.3 North Anna Power Station Revision 16 Page 17 of 27 CAUTION
- To avoid the damage caused by water hammer when placing the Reheater Warm-up System in service, the throttle valves MUST be throttled open in very SMALL increments AND sufficient time MUST be allowed between valve movements to verify temperature changes.
- Reheater outlet steam temperatures (LP Turbine inlet temperatures) MUST be closely monitored. The heat-up rate limit of 25°F in a 15-minute period MUST NOT be exceeded.
- IF a Reactor trip occurs during this Reheater warmup, THEN the Throttle Valves for the Reheater 3 inch warm-up line MUST be isolated immediately in accordance with Attachment 1, Actions for Reactor Trip During MSR Warmup, to prevent unwanted cooldown and possible reactivity transient.
NOTE: When available, Reheater Control System LP I and LP 2 are the more accurate indication of LP Turbine inlet temperatures, NOT individual Reheater outlet temperatures. Reheat Control System LP1 and LP2 temperatures should be monitored during MSR heatup. When LP I and LP 2 are NOT available, use alternate indications including computer points and the Turbine Supervisory Panel recorder. (Reference 2.4.2) 5.2.11 In the Control Room, slowly increase the Manual Valve positioner output to 100 percent and verifj the following valves indicate full open: 1-MS-FCV-104A, (VALVE NO. 1) 1-MS-FCV-104B, (VALVE NO.2) 1-MS-FCV-104C, (VALVE NO.3)
- I-MS-FCV-104D, (VALVE NO.4)
\23 I I
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 38. O4OAKI .04 03 8/NEW//H/4/3 .2/3.6/4/
Given the following conditions:
- Unit 1 was initially at 100% power.
- The team is recovering from a reactor trip due to a loss of offsite power.
- All systems functioned as designed except for AFW to C SG.
- Local actions have restored the capability to supply AFW to C SG.
The following conditions exist:
- RCS Hot-leg temperatures are 558°F and trending down slowly.
- C SG Wide-range level is 6% and trending down slowly.
Based on these conditions, 1-FR-H.5, Response to Steam Generator Low Level, directs the crew to The basis for this restriction is to prevent A. establish AFW flow to C SG at < 150 gpm ; excessive RCS cooldown B. establish AFW flow to C SG at 150 gpm thermal stress to SG components C. NOT establish AFW flow to C SG excessive RCS cooldown D. NOT establish AFW flow to C SG ; thermal stress to SG components
- a. Incorrect. Plausible since candidate would tend to want to refill the SG and may reason that reduced flow rate would be an acceptable method. Preventing excessive cooldown plausible since post trip recovery procedures temper actions to stabilize the plant, and excessive cooldown could lead to inventory issues and complicate recovery actions (150 to each SG would be the typical practiced method used by crews in this situation to ensure symmetric cooling and to aid in diagnosing any subsequent failure (since all things being the same SGs would expect to recover level at the same rate; if this were not the case it would be indicative of some other malfunction).
- b. Incorrect. First part incorrect but plausible as noted above; second part is correct as described in the WOG Background Document & Executive Volume.
- c. Incorrect. First part is correct; as aluded to above this action is conservative and ensure that the condition is evaluated, or another strategy such as cooling the RCS is used prior to re-initiating feed due to the potential consequence. Second part incorrect but plausible as discussed in Distractor a.
- d. Correct. First part is correct as discussed in Distractor c. Second part is also correct as discussed in distractor b.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Steam Line Rupture Knowledge of the operational implications of the following concepts as they apply to Steam Line Rupture: (CFR 41.8 /41.10 / 45.3) Nil ductility temperature Tier: 1 Group: 1 Technical
Reference:
EOP-FR-H.5 and WOG background document Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info: This is a resent change (EOP Rev. 2) to the procedure.
STEP DESCRIPTION TABLE FOR FR-H.5 Step 4 KNOWLEDGE: In the background document for FR-H-i, RESPONSE TO LOSS OF SECONDARY HEAT SINK, a hot, dry steam generator is defined as a SG in which the primary side of the SG tubes is above 550°F (determined by hot leg temperature readings) and the secondary side has no liquid inventory. 550°F is a temperature evaluated to be low enough that thermal stress would not lead to a failure when feedwater is established to a dry steam generator. For FR-H.5, if the SG level drops below the wide range level indication range, the steam generator is assumed to be dry. k-c 9 c In FR-H.1, a rapid restoration of feedwater may be necessary for the \f reestablishment of an adequate secondary heat sink. A rapid restoration of AFW flow is not necessary in FR-H.5 to establish level indication. Unless directed by the plant engineering staff, it is prohibited to feed a dry steam generator. A dry steam generator is defined as a steam generator with a water level below the wide range level indication. For feeding a dry steam generator, guidance is provided only in FR-H.i where a severe challenge exists on the HEAT SINK status tree. This guidance is not applicable for FR-H.5. Only one effective steam generator is necessary to maintain an adequate secondary heat sink. Following an evaluation by the plant engineering staff as part of the long term recovery actions, the affected steam generator may be refilled. This evaluation should consider steam generator materials and properties, Technical Specification considerations, etc. PLANT-SPECIFIC INFORMATION: o (S.O4) Feed flow value in plant specific units corresponding to 25 gpm. The 25 gpm value is representative of a minimum measurable feed flow to a steam generator. Plant specific values may depend upon flow instrumentation and the sensitivity of the controls on the feed flow. o (N.Oi) SG level just in the wide range, including allowances for normal channel accuracy and reference leg process errors. o (N.O2) SG level just in the wide range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors. FR-H.5 Background 15 HP-Rev. 2, 4/30/2005 HFRH5BG .doc
NORTH ANNA POWER STATION FUNCTION RESTORATION PROCEDURE NUMBER PROCEDURE TfTLE REVISION 9 1-FR-H.5 RESPONSE TO STEAM GENERATOR LOW LEVEL (WITH NO ATTACHMENTS) PAGE 1 of 3 PURPOSE To provide instructions to respond to a Iowpriority challenge to secondary system heat sink capability resulting from inventory below the narrow range in intact SGs. ENTRY CONDITIONS This procedure is entered from:
- Yellow terminus of the HEAT SINK CSF STATUS TREE.
CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 9 1-FR-H 5 RESPONSE TO STEAM GENERATOR LOW LEVEL
. PAGE 2 of 3
-_STEP ACTION! EXPECTED RESPONSE I RESPONSE NOT OBTAINED CAUTION: In order to minimize secondary inventory losses, steam releases from affected SGs should be minimized. NOTE:
- Setpoints in brackets [] are for adverse Containment atmosphere (20 psia Containment pressure or Containment Radiation has reached or exceeded I .0E5 R!hr or 70% on High Range Recorder).
In this procedure, AFFECTED SG refers to any SG in which narrow range level is below 11% [22%]
- 1. IDENTIFY AFFECTED SGs NARROW - D RETURN TO procedure and step in effect.
RANGE LEVEL LESS THAN 11% [22%1
- 2. VERIFY BLOWDOWN ISOLATION FROM D Manually close valves.
AFFECTED SGs:
- ASG:
Close 1-BD-TV-100A D
- Close 1-BD-TV-100B
- B SG:
- Close 1-BD-TV-100C 0
- Close 1-BD-TV-100D C SG:
0
- Close 1-BD-TV-100E 0
- Close 1-BD-TV-100F
NUMBER PROCEDURE TITLE REVISION 9 1-FR-H 5 RESPONSE TO STEAM GENERATOR LOW LEVEL
. PAGE 3 of 3 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED H_STEP
- 3. CHECK AFFECTED SGs - NOT FAULTED: D H affected SGs were previously identified as faulted AND are isolated, THEN return to D a) Pressure of affected SGs - GREATER procedure and step in effect.
THAN 80 PSIG D b) Pressure of affected SGs CONTROLLED
- D IF affected SGs are NOT isolated, THEN BY OPERATOR GO TO 1-E-2, FAULTED STEAM GENERATOR ISOLATION, STEP 1.
- 4. CHECK AFW FLOW TO AFFECTED SGs - D H affected SGs wide range level greater GREATER THAN 100 GPM than 13% [30%], THEN establish AFW flow as necessary to refill affected SGs.
IF affected SGs wide range level less than 13% [30%], THEN do the following: a) Do not establish AFW flow to affected SGs. b) Consult the TSC or Plant Staff to evaluate refilling the affected SGs as part of long-term plant recovery. D c) GOTOStep6. 5._ CONTINUE TO FILL AFFECTED SGs UNTIL NARROW RANGE LEVEL GREATER - THAN 11% [22%] 6._ RETURN TO PROCEDURE AND STEP IN EFFECT
- END -
STEP DESCRIPTION TABLE FOR FR-H.5 Step 4 STEP: Check AFW Flow To Affected SG(s) - GREATER THAN (S.O4) GPM PURPOSE: To determine if AFW flow exists to the affected SG BASIS: If feed flow to a SG is isolated and the SG is allowed to dry out, subsequent reinitiation of feed flow to the SG could create significant thermal stress conditions on SG components. Maintaining a minimum verifiable feed flow (i.e., (S.04) gpm) to the SG allows the components to remain in a wet condition, thereby minimizing any thermal shock effects if feed flow is increased. In addition, if SG wide range indication shows that an appreciable water mass exists (i.e., greater than (N.O1)% [(N.O2)% for adverse containment], in the SG, then reestablishment of AFW should not cause significant thermal stresses. However, if SG wide range level is less than (N.O1)% [(N.O2)% for adverse containment], then the SG may have dried out. For this condition, AFW feed flow should not be established until the plant engineering staff performs an evaluation as part of the long term recovery operations. ACTIONS: o Determine if AFW flow to affected SG(s) is greater than (S.O4) gpm o Determine if affected SG wide range level is greater than (N.Ol)% [(N.O2)% for adverse containment] o Establish AFW flow as necessary to refill affected SG(s) INSTRUMENTATION: o AFW flow indication o SG wide range level indication CONTROL/EQUIPMENT: AFW flow control valve switches FR-H.5 Background 14 HP-Rev. 2, 4/30/2005 HFRH5BG .doc
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 39. 045K3 .01 03 9/NEW//H/4/2.9/3 .2/4/
Given the following conditions:
- Unit 1 was initially stable at 50% power.
- RCS TAVG & TREF were initially matched and on program.
An event occurs, which causes rods to begin stepping IN. The crew observes the following indications:
- TAVG is 568°F.
- TREE is 560° F.
- Power Range Channel N-44 is 48%.
Which ONE of the following identifies the correct response to this event AND includes the reason for this response? A. Place rods in MANUAL because they are moving in response to a N-44 failure. B. Place rods in MANUAL because they are moving in response to a TREE failure. C. Leave rods in AUTO because they are moving in response to an increase in steam demand. D. Leave rods in AUTO because they are moving in response to a decrease in steam demand.
- a. Incorrect. Plausible since the operator may not analyze the plant conditions for the given power level and erroneously focus on N-44 solely as the cause. However, the given 48% power indication is a consequence of the inward rod motion, not the cause.
- b. Incorrect. Plausible again if the operator isolates on the disparity between Tavg and Tref as the cause they may erroneously select this distractor. This is another case where the Tref indication is a consequence of the event, not the cause.
- c. Incorrect. Although leaving rods in auto is appropriate the reason for that response is not. The operator may conclude that Tref is lower because steam flow is being robbed from the turbine, and while this could be argued it does not explain the Tave value and thus does not constitute the correct reason.
- d. Correct. Leaving rods in auto is the correct response. As previously discussed the reduction in N-44 level is an expected consequence of the inward rod motion. Tref is a function of first stage pressure, thus an event such as a turbine governor valve failing closed would cause a reduction in Tref; the Tave increase is a consequence of the reduction in heat removal caused by the governor valve failing closed.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Main Turbine Generator (MT/G) System Knowledge of the effect that a loss or malfunction of the MT/G system will have on the following: (CFR: 41.7 /45.6) Remainder of the plant Tier: 2 Group: 2 Technical
Reference:
AR B-A7, 1-AP-1.1, 1-AP-4.3, 1-AP-3 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
VIRGINIA POWER 1-EI-CB-21B ANNUNCIATOR A7 1-AR B A7 NORTH ANNA POWER STATION REV. 4 APPROVAL: ON FILE Effective Date:08/04/09 MEDIAN/HI TAVG
< > TREF +/- 5°F DEVIATION 1.0 Probable Cause 1.1 Step change in indicated Tavg or Tref 1.2 Load or power transient 1.3 Failure of 1st stage pressure channel 1.4 Failure of Tavg Control circuitry 1.5 Cooldown in progress 1.6 Failure to operate rods or borate or dilute appropriately.
2.0 Operator Action ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 VERIFY 1st STAGE PRESSURE - GO TO l-AP-3, Loss of Vital NORMAL Instrumentation 2.2 VERIFY l-RC-TI-l4O8A Do the following: MEDIAN/HI TAVG - NORMAL a) Place Control Rods in MANUAL b) Place PRZR level control in MANUAL c) Place Steam Dumps in STEAM PRESSURE mode as follows:
- 1) Place both STEAM DUMP INTLK switches to Ic OF F / RESET 1R 2) Place STEAM DUMP CONTROLLER to MANUAL
- 3) Place MODE SELECTOR switch to STEAM PRESS f9 4) Ensure Steam Dump demand is ZERO
- 5) Return STEAM DUMP CONTROLLER to AUTO
- 6) Verify Steam Dump demand is ZERO I c 7) Place both STEAM DUMP INTLK switches to ON d) Submit a Work Request to repair Median/Hi TAVG control circuit e) IF the Reactor is critical and RCS Tavg is < 547°F with the Tavg - Tref Deviation alarm NOT reset, THEN log Tavg every 30 minutes
2.3 CHECK UNIT 1 - IN MODE 1 OR 2 IF controlled cooldown in progress, THEN return to procedure and step in effect. Use the following to control or minimize transients as required:
- Steam load adjustments
- Feedwater adjustments
- Steam trap alignment
- MSR alignment
- SG Blowdown adjustments IF Tavg continues to decrease after steam loads, SG blowdown flow, and feedwater flows have been minimized, THEN close MSTV5 and MSTV Bypass valves.
Return to procedure in effect. Note: Adding positive reactivity is NOT an appropriate method to address an unstable plant condition. 2.4 VERIFY UNIT CONDITIONS - STABLE IF unable to control reactor OR UNDER CONTROL power, THEN go to l-AP-38, a) Reactor Power EXCESSIVE LOAD INCREASE. b) Secondary load Otherwise, use the following to
- Main Turbine control or minimize transients
- Steam Dumps as required
- PORV5
- Control Rod Manipulations
- Aux Steam
- Boration/Dilution c) Steam Generator Levels
- Steam Load adjustments
- Feedwater adjustments 2.5 Check Tavg - LESS THAN 547°F Return to procedure in effect.
2.6 Initiate l-LOG-l4 to record Tavg every 30 minutes while the annunciator is in alarm and Tavg is less than 547°F. 2.7 CHECK EACH RCS LOOP TAVG - Do one of the following within GREATER THAN 541°F 30 minutes in accordance with ITS 3.4.2, Condition A:
- Restore TAVG to greater than 541°F.
- Be in Mode 2 with Keff <1.0.
3.0 References 3.1 Westinghouse Logic 5655D33 3.2 ll7l5-ESK-1OAAR 3.3 ll7ls-FM-93A 3.4 Unit 1 Loop Book, page RC-88 3.5 DCP 89-40-1 3.6 ITS 3.4.2 3.7 CTS 02-95-2127-003, Tech Spec Review 4.0 Actuation 4.1 1-RC--TC-1408D
Doinien NORTH ANNA POWER STATION ABNORMAL PROCEDURE NUMBER PROCEDURE TITLE REVISION 9 1-AP-1.i CONTINUOUS UNCONTROLLED ROD MOTION (WITH TWO ATTACHMENTS) PAGE 1 of 5 PURPOSE To provide instructions to follow in the event of continuous uncontrolled rod motion. ENTRY CONDITIONS This procedure is entered when any of the following conditions exists:
- Annunciator Panel A H-i, ROD BANK A LO/LO-LO LIMIT, is LIT, or
- Annunciator Panel A H-2, ROD BANK B LO/LO-LO LIMIT, is LIT, or
- Annunciator Panel A H-3, ROD BANK C LO/LO-LO LIMIT, is LIT, or
- Annunciator Panel A H-4, ROD BANK D LO/LO-LO LIMIT, is LIT, or
- Annunciator Panel B A-7, MEDIAN/HI TAVG <>TREF DEVIATION, is LIT, or
- Undesired Tavg change with rod motion occurring, or
- Undesired power change with rod motion occurring.
(&. 2 CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 9 1-AP-1 1 CONTINUOUS UNCONTROLLED ROD MOTION PAGE _ 2 of 5 -_STEP_. ACflON/ EXPECTED RESPONSE I RESPONSE NOT OBTAINED 1] PUTCONTROLROD BANK i i_v SELECTOR SWITCH TO ç cJ 4 ( e c 2 j_ VERIFY ROD MOTION - D GO TO 1-E-0, REACTOR TRIP OR SAFETY STOPPED INJECTION.
- 3. VERIFY 1-RC-TI-1408A MEDIAN/HI D Initiate actions of Annunciator Panel B A-7 TAVG NORMAL
- MEDIAN/HI TAVG <>TREF DEVIATION, while continuing with this procedure.
NOTE:
- When the Unit is in Mode 1 or Mode 2 with Keff greater than or equal to 1.0, then shutdown margin is restored when the control rods are above the LO-LO insertion limit.
Continuous control rod motion should be avoided to prevent excessive reactivity addition.
*4 MAINTAIN THE FOLLOWING USING CONTROL RODS AND BORATION:
D a) Rod Bank LO/LO-LO Limit a) Do the following to ensure Tech Spec 3.1.1 Annunciators NOT LIT and 3.1 .6 compliance: D 1) Immediately initiate boration until the required shutdown margin is restored. E1 2) Restore Control Rods to above the Insertion Limits within 30 minutes. D b) A.F.D. Monitor EJ b) IF alarm is valid, THEN restore AFD within limits Annunciator Panel A H-7 - OR reduce thermal power to <50% RTP within NOT LIT 30 minutes in accordance with Tech Spec 3.2.3.
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-1.1 2 TREE PROGRAM .338 °EI% POWER REVISION PAGE 9 lofi 585 I j r
- I*
j I 1H I q II
. 570 565-I Tave JMAX f +÷-
ti: j ;v j: r LtI/f f;J./.l jTT:ijT/r7ii 1Th1T FflT HTR I I4* IJJ1J I Lii I-i LiLLLLLL Tave iFH I I I I I LLLHJ I I I j I I IJLIJ ILL I I 555 rt_Lj I I UM I LW L
! ii .II!
Power Level Vi -2006-09-28 NOTE: Tave should be maintained with 1 .5°F of Tref.
Doniuion NORTH ANNA POWER STATION ABNORMAL PROCEDURE NUMBER PROCEDURE TITLE REVISION 19 1 -AP-4.3 MALFUNCTION OF NUCLEAR INSTRUMENTATION (POWER RANGE) PAGE 1 of 11 (WITH EIGHT ATTACHMENTS) PURPOSE To provide instructions to follow in the event of a malfunction in the Power Range Instrumentation channels. ENTRY CONDITIONS This procedure is entered as directed by the Nuclear Instrument Annunciator Response or when any of the following conditions exist:
- Erratic or erroneous indication on Power Range instruments, or
- Loss of indication on Power Range instruments, or
- Invalid trip signals, or
- Invalid rod stop.
CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 19 1 -AP-4.3 MALFUNCTION OF NUCLEAR INSTRUMENTATION (POWER RANGE) PAGE 2 of 11 H_STEP_jj ACTION/EXPECTEDRESPONSE I RESPONSENOTOBTAINED I cJ 1 j- STOP POWER INCREASE 2 ]_ VERIFY N-44 NOT FAILED D Place Rod Control and Main Feed Reg Bypass Valves in MANUAL. CAUTION:
- Adjusting N-44 current with the control rods or Main Feed Reg Bypass Valves in automatic may cause the control rods or Main Feed Reg Bypass Valves to move.
- If more than one Power Range instrument is inoperable, then placing a second channel in the tripped condition will result in a Reactor trip.
NOTE: If more than one Power Range Instruments is inoperable, then annunciator Panel A D-4, is also inoperable.
- 3. VERIFY THREE POWER RANGE Do the following:
INSTRUMENTS OPERABLE a) Perform required actions of Technical Specification 3.0.3. b) Do the following: l
- Refer to Tech Spec 3.3.1, Table 3.3.1-i, Functions 2, 3, and 6
- Refer to the following Tech Spec 3.3.1, Table 3.3.1-1 Functions within one hour:
- Function 18.b (P-7, Panel P G-2)
- Function 18.c (P-8, Panel P F-i)
- Function 18.d (P-b, Panel P D-2)
(STEP 3 CONTINUED ON NEXT PAGE)
p Dernnoi NORTH ANNA POWER STATION ABNORMAL PROCEDURE NUMBER PROCEDURE TITLE REVISION 24 1 -AP-3 LOSS OF VITAL INSTRUMENTATION (WITH TWO ATTACHMENTS) PAGE 1 of 19 PURPOSE To provide instructions to follow in the event of a loss of vital instrumentation. ENTRY CONDITIONS This procedure is entered when a faulty indication occurs on any of the following vital instrumentation channels:
- Reactor Coolant Flow, or
- Pressurizer Level, or
- Pressurizer Pressure Protection, or
- DELTA T/TAVE Protection, or
- Containment Pressure Protection, or
- RWST Level, or
- Steam Generator Level, or
- Turbine Stop Valves Indication, or
- Turbine First Stage Impulse Pressure, or
- Turbine Auto Stop Oil Low Pressure Trip Signal, or
- Steam Flow, or
- Feed Flow, or
- Steam Pressure, or
- Station Service Bus Undervoltage, or
- Station Service Bus Undertrequency.
CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 24 1 -AP-3 LOSS OF VITAL INSTRUMENTATION PAGE 2 of 19 -_STEP ACTION/EXPECTEDRESPONSE I I RESPONSENOTOBTAINED 1] VERIFY REDUNDANT Q IF unable to determine Reactor is in a safe INSTRUMENT CHANNEL operating condition, THEN GO TO 1-E-O, INDICATION NORMAL REACTOR TRIP OR SAFETY INJECTION. 2] VERIFY STEAM GENERATOR Do the following: LEVEL CONTROLLING CHANNELS NORMAL:
- a) Place the associated valves in MANUAL:
D
- Steam Flow
- Main Feed Reg Valves D
- Feed Flow D
- Main Feed Reg Bypass Valves D
- Steam Generator Level Ch Ill b) Control Steam Generator level.
D
- Steam Pressure 3 j VERIFY TURBINE FIRST STAGE IF the controlling channel failed, THEN do the PRESSURE INDICATIONS - following:
NORMAL icR 1
- Place Control Rod Mode Selector switch in MANUAL.
- Manually control SG level on program, as desired.
4 1 VERIFY PRESSURIZER LEVEL IF any selected channel failed, THEN do the INDICATIONS NORMAL
- following:
a) Place 1-CH-FCV-1122, Charging Flow Control Valve, in MANUAL. D b) Control Pressurizer level at program.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 40. 055EK1.02 040/BANK/PRAIRIE ISLAND: 2005/L/3/4.1/4.4/6/
Given the following conditions:
- 1-ECA-0.0, Loss of All AC Power, is in progress.
- The crew is performing step 22, Depressurize all Intact SGs to 290 psig.
- The BOP mistakenly did NOT stop depressurizing SGs at 290 psig.
- Pressure in all SGs reaches 190 psig before the depressurization is stopped.
What is the potential operational implication that could result from the excessive SG depressurization? A. Nitrogen injection from the accumulators may occur, causing natural circulation flow in the RCS to be interrupted. B. Transition to 1-FR-P.1, Response to Imminent Pressurized Thermal Shock, could be required due to the excessive RCS cooldown. C. Voiding may occur in the reactor vessel, causing the upper portion of the core to become uncovered and potentially causing core damage. D. An undesired automatic Safety Injection signal may occur, complicating other recovery actions that are in progress.
- a. Correct. Answer was abbreviated from the background document which specifically states Maintaining steam generator pressures above a value that pre vents introduction of a significant volume of nitrogen into the RCS ensures that accumulator nitrogen will not impede natural circulation .
- b. Incorrect. Plausible because ECA-0.0 is written implicitly to avoid challenges to CSFs. However even if the Subcriticality CSF was challenged, action would not be taken because there is no power (FR5 are monitored for information only and not implemented until specifically directed when in ECA-0 series procedures).
- c. Incorrect. Plausible since there is a concern for upper head voiding; however this again is a case where some overshoot of SG pressure is anticipated, and core damage is not an expected consequence for the given conditions.
- d. Incorrect. Plausible since, if actuated, SI is reset to allow component manipulations as required by the procedure. In this case an SI signal results as a consequence of the higher order requirement to depressurize SGs. The candidate who does not realize this consequence may assume that it would only result from overshoot and erroneously conclude that this distractor is a legitimate concern.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Loss of Offsite and Onsite Power (Station Blackout) Knowledge of the operational implications of the following concepts as they apply to the Station Blackout: (CFR 41.8 / 41.10 / 45.3) Natural circulation cooling Tier: 1 Group: 1 Technical
Reference:
EOP-ECA-0.0 AND WOG Background Document Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info: Distractor D modified EOP change 2/11/2010 changed this number from 145 psig to 290 psig and caution changed from 120 psig to 190 psig.
NUMBER PROCEDURE TITLE REVISION 23 1 -ECA-O.O LOSS OF ALL AC POWER PAGE 14 of 22 STEP ACTION/EXPECTED RESPONSE J RESPONSE NOTOBTAINED I CAUTION:
- To prevent injection of SI Accumulator nitrogen n the RCS, SO pressure should be maintained greater than 190 psig.
To prevent loss of heat sink, narrow range level should be maintained greater than 11% [22%] in at least one intact SG. If level cannot be maintained, then depressurization should be stopped until level is restored in at least one intact SG. a CI- i-zAI q I H-NOTE:
- Depressurization of SGs will result in SI actuation.
- To minimize RCS inventory loss, the SGs should be depressurized at the maximum controllable rate, not to exceed an RCS cooldown of 100°F/hr.
9 ,
- Depressurization of SGs should be continued even if PRZR level is lost or if Reactor Vessel Upper Head voiding occurs.
- Communications should be established between the Control Room and the Main Steam Valve House during local operation of the SG PORVs using sound powered phones or radios.
22._ DEPRESSURIZE ALL INTACT SGs TO 290 PSIG: D a) Check narrow range level GREATER
- a) Do the following:
THAN 11%[22%] IN AT LEAST ONE
- 1) Maintain maximum AFW flow until INTACT SG narrow range level is greater than 11% [22%] in at least one intact SO.
- 2) WHEN narrow range level is greater than 11% [22%] in at least one intact SO, THEN RETURN TO Step 22.
Continue with Step 23. (STEP 22 CONTINUED ON NEXT PAGE)
STEP DESCRIPTION TABLE FOR ECA-O.O Step 16 - CAUTION 1 CAUTION: SG pressures should not be decreased to less than (0.07) psig to prevent injection of accumulator nitrogen into the RCS. FZ eAr PURPOSE: To alert the operator that steam generator pressures must be maintained above the specified limit BASIS: Steam generators should be depressurized to maximize delivery (into the RCS) of the water contained in the SI accumulators while minimizing delivery of nitrogen. Maintaining steam generator pressures above a value that prevents introduction of a significant volume of nitrogen into the RCS ensures that accumulator nitrogen will not impede natural circulation. A steam generator pressure limit is set to preclude significant nitrogen injection into the RCS. To determine the steam generator pressure limit, an ideal gas expansion calculation should be performed based on nominal plant
) initial nitrogen specific val ues for initial accumul ator tank pressure (P 1 ,
gas volume (V), and final nitrogen 1 gas volume ) 2 (V. The final nitrogen gas volume should be equivalent to the total accumulator tank volume. The RCS pressure at empty tank conditions (P ) is determined from: 2 DI( = DI( where y = 1.25 for ideal gas expansion. The steam generator pressure limit is then determined by subtracting the RCS to SG delta P from P . The RCS to SG 2 delta P should be calculated as described in the RCP TRIP/RESTART section in the Generic Issues of the Executive Volume. Instrument uncertainties are not included in the determination of the steam generator pressure limit to preclude a bias toward either having more accumulator water injected into the RCS or having less nitrogen injected into the RCS. ACTIONS: Determine if SG pressures are greater than (0.07) psig INSTRUMENTATION: SG pressure indication for each SG ECA-0.0 Background 115 HP-Rev. 2, 4/30/2005 HECAOOBG.doc
STEP DESCRIPTION TABLE FOR ECA-O.O Step 16 - CAUTION 1 CONTROL/EQUIPMENT: N/A KNOWLEDGE: N/A PLANT-SPECIFIC INFORMATION: (0.07) Minimum SG pressure which prevents injection of accumulator nitrogen into the RCS. Refer to background document for guideline ECA-O.O. ECA-O.O Background 116 HP-Rev. 2, 4/30/2005 HECAOOBG.doc
STEP DESCRIPTION TABLE FOR ECA-O.O Step 16 - CAUTION 2 CAUTION: SG narrow range level should be maintained greater than (M.02)% [(M.O3)% for adverse containment] in at least one intact SG. If level cannot be maintained, SG depressurization should be stopped until level is restored in at least one SG. PURPOSE: To inform the operator of the importance of maintaining at least one intact SG narrow range level above the top of the U-tubes during depressurization BASIS: During the rapid depressurization performed in Step 16, SG level could drop out of the narrow range resulting in a loss of adequate heat sink. If this situation occurs, the depressurization should be stopped and AFW flow reestablished until SG narrow range level is increased to greater than (M.O2)% [(M.O3)% for adverse containment]. The analysis basis for ECA-O.O requires that the level in at least one intact steam generator (SG) be above the top of the SG U-tubes to ensure that sufficient heat transfer capability exists to remove heat from the RCS via either natural circulation or reflux boiling after the RCS saturates. This is accomplished in the guideline by Step 13 (which requires maximum AFW flow to be maintained to the intact SGs until level in at least one intact SG is in the narrow range) and the second caution before Step 16 (which requires that level in at least one intact SG be maintained in the narrow range during SG depressurization). Once these conditions are met, Step 16 directs the operator to dump steam (depressurize intact SGs) at a maximum rate to reduce RCS temperature and pressure (which in turn will reduce the rate of RCS inventory loss through the RCP seals). Step 16 is structured to provide flexibility in how and at what speed the intact SGs are depressurized. As discussed in the BASIS Section for Step 16, maximum rate means in a controlled manner subject to plant specific constraints. Depending on plant specific design considerations, the depressurization may be performed using one, more than one, or all intact SGs. This flexibility is intended to permit utilities to structure their EOP step (and train their operators) in the way that best suits their plant design. ECA-O.O Background 117 HP-Rev. 2, 4/30/2005 HECAOOBG .doc
STEP DESCRIPTION TABLE FOR ECA-O.O Step 16 - CAUTION 2 As discussed in the BASIS Section for Step 16, the ERG reference plant has air operated atmospheric dump valves (one valve per steamline) which have dc control power and pneumatic power (i.e., either air reservoirs or nitrogen bottles) available following the loss of all ac power. For a plant like the reference plant that has the ability to control these valves on all intact SGs from the control room, the preferred method to depressurize the intact SGs is to uniformly release steam from all intact SGs at a controlled rate that will not cause the level in the SGs to drop out of the narrow range. The BASIS Section for Step 16 also notes that some plants may not have the ability to open and control the SG PORVs from the control room and may have to rely on local manual actions to depressurize the intact SGs. Also, some plants (including the ERG reference plant) have only one SG PORV per steamline. To accommodate local operator actions or the inability to open or control the PORV on each intact SG steamline, the loss of all ac power analysis basis and the structure of Step 16 permits SG depressurization using one, more than one, or all intact SGs. For example, it is acceptable to keep one intact SG isolated (with level in the narrow range) while depressurizing the other (i.e., one or more than one) intact SGs. Depending on plant specific considerations (e.g., local actions needed to open and control the intact SG PORV5), depressurization in this manner may avoid the potential to lose level in all intact SGs which would require the depressurization to be stopped until level could be restored in at least one intact SG. Starting and stopping the depressurization in this manner may complicate local manual actions and potentially hinder the overall depressurization. If the depressurization is performed with one intact SG isolated, the isolated intact SG should eventually be unisolated and depressurized once the concern for losing level in all intact SGs no longer exists. ACTIONS: N/A INSTRUMENTATION: N/A CONTROL/EQUIPMENT: N/A ECA-O.O Background 118 HP-Rev. 2, 4/30/2005 HECAOOBG.doc
STEP DESCRIPTION TABLE FOR ECA-O.O Step 16 - CAUTION 2 KNOWLEDGE: N/A PLANT-SPECIFIC INFORMATION: o (M.02) SG level just in the narrow range, including allowances for normal channel accuracy and reference leg process errors. o (M.03) SG level just in the narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50%. ECA-O.O Background 119 HP-Rev. 2, 4/30/2005 HECAOOBG .doc
STEP DESCRIPTION TABLE FOR ECA-O.O Step 16 - NOTE 1 NOTE: The SGs should be depressurized at maximum rate to minimize RCS inventory loss. PURPOSE: To inform the operator of the desired rate for depressurization of steam generators BASIS: The intact steam generators should be depressurized as quickly as possible, to minimize RCS inventory loss, but within the constraint of controllability. Controllability is required to ensure that steam generator pressures do not undershoot the specified limit. For the reference plant, the operator can control the secondary depressurization from the control room. In this case, maximum rate means steam generator PORVs full open. For plants that must control the secondary depressurization by local actions, maximum rate must be determined by the control room and local operators based on plant conditions and available communications. A slower rate is acceptable for locally controlled secondary depressurization. See Subsection 2.3. ACTIONS: N/A INSTRUMENTATION: N/A CONTROL/EQUIPMENT: N/A KNOWLEDGE: SG depressurization should proceed as quickly as possible and should not be limited by the Technical Specification RCS cooldown limit of 100°F/hr. PLANT-SPECIFIC INFORMATION: N/A ECA-O.0 Background 120 HP-Rev. 2, 4/30/2005 HECAOOBG .doc
STEP DESCRIPTION TABLE FOR ECA-O.O Step 16 - NOTE 2 NOTE: PRZR level may be lost and reactor vessel upper head voiding may occur due to depressurization of SGs. Depressurization should not be stopped to prevent these occurrences. PURPOSE: To inform the operator of possible reactor vessel upper head voiding during steam generator depressurization BASIS: Loss of pressurizer level and reactor vessel upper head voiding may result from the rapid depressurization of the intact steam generators. Such a condition is anticipated and should not interfere with operator actions in Step 16 to depressurize the steam generators to reduce RCS pressure and temperature and to minimize RCS inventory loss out of the RCP seals. ACTIONS: N/A INSTRUMENTATION: N/A CONTROL/EQUIPMENT: N/A KNOWLEDGE: N/A PLANT-SPECIFIC INFORMATION: N/A ECA-O.O Background 121 HP-Rev. 2, 4/30/2005 HECAOOBG .doc
STEP DESCRIPTION TABLE FOR ECA-O.O Step 16 STEP: Depressurize Intact SGs To (0.08) PSIG PURPOSE: To depressurize the intact steam generators BASIS: Step 16 depressurizes the intact SGs, thereby reducing RCS temperature and pressure to reduce RCP seal leakage and minimize RCS inventory loss. The advantages to performing this action, as well as restrictions that apply during the action, are detailed in Subsection 2.3. During SG depressurization, SG level must be maintained above the top of the SG U-tubes in at least one SG. Maintaining the U-tubes covered in at least one SG will ensure that sufficient heat transfer capability exists to remove heat from the RCS via either natural circulation or reflux boiling after the RCS saturates. Step 16a requires that SG level be in the narrow range in at least one SG before SG depressurization is initiated in Step 16b. If level is not in the narrow range in at least one SG, RNO 16a instructs the operator to maintain maximum AFW flow until narrow range level is established in one SG. When narrow range level is established, SG depressurization can be started or continued via Step 16b. Step 16b instructs the operator to reduce SG pressures by depressurizing the intact SGs. Depressurization should be accomplished by opening the PORVs on the intact SGs to establish a maximum steam dump rate, consistent with plant specific constraints. The step is structured assuming that the operator can open and control SG PORVs from the control room. This structure assumes that the PORVs are air-operated and have dc control power and pneumatic power (i.e., either air reservoirs or nitrogen bottles) available. Some plants may not have the capability to open the SG PORVs from the control room. These plants should evaluate their capability to accomplish this step locally via PORV handwheels. Such an evaluation should consider accessibility and communications necessary to accomplish local PORV operation. Once depressurization is initiated, maintenance of a specified rate is not critical. The depressurization rate should be sufficiently fast to expeditiously reduce SG pressures, but not so fast that SG pressures cannot be controlled. It is important that the depressurization not reduce SG pressures in an uncontrolled manner that undershoots the pressure limit, thus permitting potential introduction of nitrogen from the accumulators into the RCS. ECA-O.O Background 122 HP-Rev. 2, 4/30/2005 HECAOOBG .doc
STEP DESCRIPTION TABLE FOR ECA-O.O Step 16 During SG depressurization, AFW flow may have to be increased to maintain the required SG narrow range level. Control of AFW flow will have to be performed from the control room or locally depending on plant specific design. Full AFW flow should be established to any SG in which level drops out of the narrow, range. ç r (. S RCS cold leg temperatures should be monitored during SG depressurization to ensure that the depressurization does not impose a challenge to the Integrity Critical Safety Function. This check is included in Step 16c since guideline ECA-O.O has priority over the Function Restoration Guidelines and the operator is instructed to not implement a Function Restoration Guideline even if a Critical Safety Function challenge is detected by the Critical Safety Function Status Trees. Consequently, Step 16c implicitly protects the Integrity Critical Safety Function. The SG depressurization should not result in a challenge to the Integrity Critical Safety Function since the resultant RCS cold leg temperatures should not approach the temperature limit (i.e., T2 temperature) at which a challenge will exist. Once the target SG pressure is reached, the SG PORVs and AFW flow should be controlled to maintain SG pressure at the target value until ac power is restored. The target SG pressure for Step 16 should ensure that RCS pressure is above the minimum pressure to preclude injection of accumulator nitrogen into the RCS. The target SG pressure should be based on the nominal SG pressure to preclude nitrogen addition, plus margin for controllability (e.g., 100 psi). To determine the steam generator pressure limit, an ideal gas expansion calculation should be performed based on nominal plant specific values for initial accumul ator tanks pressure (P ), initial nitrogen gas vol ume (V 1 ), and 1 final nitrogen gas volume (V). The final nitrogen gas volume should be 2 equivalent to the total accumulator tank volume. The RCS pressure at empty tank conditions (P) is determined from: 2 pv 11 = PV 22 where y = 1.25 for ideal gas expansion. The steam generator pressure limit is then determined by subtracting the RCS to SG delta p from P 2 and adding the margin to controllability. The RCS to SG delta p should be calculated as described in the RCP TRIP/RESTART section in the Generic Issues of the Executive Volume. Instrument uncertainties are not included in the determination of the steam generator pressure limit to preclude a bias toward either having more accumulator water injected into the RCS or having less nitrogen injected into the RCS. ECA-0.0 Background 123 HP-Rev. 2, 4/30/2005 HECAOOBG .doc
STEP DESCRIPTION TABLE FOR ECA-O.O Step 16 In addition to the accumulator nitrogen limitation on SG depressurization, Subsection 2.3 also discusses the core criticality concern that exists when the RCS is cooled down. This concern was evaluated for the reference plant for various fuel burnups, assuming equilibrium xenon, all rods inserted and no addition of borated water (which will occur when RCS pressure is decreased below approximately 650 psig). The evaluation showed that only at the end of core life did the criticality concern (418°F for the reference plant) become more limiting than the accumulator nitrogen concern (410°F saturation for the reference plant). For the assumed conditions, the accumulator nitrogen concern dominated the criticality concern. Consequently, Step 16 is structured to explicitly address the accumulator nitrogen concern. For the majority of circumstances, this will cover the criticality concern. Step 17 explicitly addresses the criticality concern by terminating the SG depressurization if core shutdown is lost. ACTIONS: o Determine if SG pressures are at (0.08) psig o Determine if SG levels are greater than (M.02)% [(M.O3) % for adverse contai nment] o Determine if RCS cold leg temperatures are greater than (I.O2)°F [(I.03)°F for adverse containment] o Monitor SG pressures o Monitor SG levels o Monitor AFW flow o Control SG PORVs to control SG pressures o Control AFW flow to control SG levels INSTRUMENTATION: o AFW flow indication for each SG o RCS cold leg temperature indication o SG pressure indication for each SG o Indication of position for SG PORVs o SG narrow range level indication for each SG o Plant specific instrumentation for control of AFW flow CONTROL/EQUIPMENT: o Controls for SG PORVs o Plant-specific controls for control of AFW flow ECA-0.O Background 124 HP-Rev. 2, 4/30/2005 HECAOOBG .doc
STEP DESCRIPTION TABLE FOR ECA-O.O Step 16 KNOWLEDGE: o PTS concerns o RCP seal integrity concerns o Relationship of RCP seal leakage to RCS pressure o Basis for SG pressure limit on SG depressurization o Basis for maintaining SG level above U-tubes PLANT-SPECIFIC INFORMATION: o Evaluate capability to control SG PORVs following loss of ac power o (M.02) SG level just in the narrow range, including allowances for normal channel accuracy and reference leg process errors. o (M.03) SG level just in the narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50%. o (0.08) Minimum SG pressure which prevents injection of accumulator nitrogen into the RCS, plus a margin for controllability. Value should be based on Footnote (0.07) which is the minimum SG pressure to which prevents injection of accumulator nitrogen into the RCS, plus a margin for controllability to ensure that the SG depressurization limit is not violated. o (1.02) RCS cold leg temperature corresponding to temperature T2, including allowances for normal channel accuracy. Refer to background document for status tree F-O.4. o (1.03) RCS cold leg temperature corresponding to temperature T2, including allowances for normal channel accuracy and post accident transmitter errors. Refer to background document for status tree F-O.4. ECA-O.0 Background 125 HP-Rev. 2, 4/30/2005 HECAOOBG .doc
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 41. 055G2.4.3 041/NEW//L/4/3.7/3.9/4/
Which ONE of the following identifies the post-accident instrument required by TR 3.3.9, Regulatory Guide (RG) 1.97 Instrumentation, AND includes the TR 3.3.9 Mode(s) of Applicability? A. 1-MS-RM-170, A Steamline Radiation Monitor; Mode 1 ONLY. B. 1-MS-RM-170, A Steamline Radiation Monitor; Modes 1, 2, or 3. C. 1-SV-RM-121, Condenser Air Ejector Radiation Monitor; Mode 1 ONLY. D. 1-SV-RM-121, Condenser Air Ejector Radiation Monitor; Modes 1,2, or 3.
- a. Incorrect. First part correct as described below. Second part is incorrect, but plausible since for other functions (TR 3.4.5) it would be required only in Mode 1 &2. TR bases discusses power levels of 25%
and 30% for declaring Steam Line and air ejector monitors functional based on sufficient isotope concentrations, so the candidate who lacks detailed knolwedge could easily conclude that since they cannnot be declared functional until well into Mode 1 operation that they would thus only be required in Model.
- b. Correct. Table 3.3.9-1 lists this as a Cat 2, type E variable. Modes are correct.
- c. Incorrect. First part is incorrect but plausible since it is required by TR (3.4.5), but not this spec as it is not qualified as RG 1.97 instrumentation. Second part is also incorrect, but plausible since as noted previously if the candidate does not have detailed knowledge of the TRM they could erroneously deduce that it would make sense to only require it for times when the Unit is on-line. If they confuse the TR 3.3.9 and TR 3.4.5 requirements they may select this distractor. Also as noted above there is a threshold power level that must be met for this instrument to be declared functional.
- d. Incorrect. First part is incorrect but plausible as discussed in Distractor c; Second part is correct as discussed in answer b.
Condenser Air Removal System (CARS) Ability to identify post-accident instrumentation. (CFR: 41.6 / 45.4) Tier: 2 Group: 2 Technical
Reference:
TR 3.3.9, TR 3.4.5 and bases Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
NUCLEAR DESIGN INFORMATION PORTAL Regulatory Guide 1.97 Instrumentation 3.3.9 Table 3.3.9-1 (Sheet 11 of 13) Unit 1, Regulatory Guide 1.97Post Accident Monitoring Equipment Indication Channel Variable Cat Type No. of Action for Action for Trending Loss of Name Channels One Channel Zero Channels Required Trending Required for Functional Functional Only 1.97 per unit 1-RMLOOP-RMS- 156 Radioactive 3 E02 1 channel N/A B Yes B Exposure Rate 1-RM-RR-150 (Sample Room) or PCs 1-RM-LOOP-RMS-157 Radioactive 3 E02 1 channel N/A B Yes B Exposure Rate 1RM-RR- 150 (Main or n ont. Rm) PCS 1-RM-LOOP-R-MS17O 1-RM-LOOP-R-MS171 1-RM-LOOP-R-MS172 1-RM-LOOP-R-MS176 I Steam Generator Relief Vent Effluent Monitor TDAFW Pump Exhaust Effl uent Monitor 2 2 E08 EO8 1 channel per SG 1 channel N/A N/A TRM 3.3.7 TRM 3.3.7 Yes Trended on PCS Yes Trended on PCS A A 1-RMLOOP-R-RMS165 Contai nment 1 A-19 2 channels IS 3.3.3 IS 3.3.3 Yes IS 3.3.3 Area Radiation 1-RM-RR-165 1-RM-LOOP-R-RMS166 1-RM-RR-166 (High Range) or PCS 1-RM-LOOP-R-SW124 RS Heat 2 C14 1 channel per N/A A Yes A Exchanger SW RSHX 1-RM-RR100 1-RM-LOOP-R-SW125 or Outlet 1-RM-LOOP-R-SW126 Radiation PCS 1RMLOOPR-SW127 Moni tors 1-RM-LOOPR-VG-179-1 Ventilation 2 E07 Upper and N/A IRM 3.3.7 Yes A Vent A Lower Range 1-RM-RR179 1-RMLOOP-R-VG-179-2 Channels or Effl uent Moni tor 1-RMLOOP-R-VG- 180-1 Ventilation 2 E-07 Upper and N/A TRM 3.3.7 Yes A Vent B Lower Range 1-RM-RR-180 1-RM-LOOPR-VG- 1802 Channels or Effluent Moni tor 1-RPI - Rod Bottom Control Rod 3 802 1 channel per N/A B No N/A Indication Position Rod. 1RSLOOPL151A Containment WR 1 A-18 2 channels IS 3.3.3 IS 3.3.3 Yes IS 3.3.3 Sump Level 1RS-LR-151B 1RSLOOPL151B or PCS 1RSLOOPP152A Inside 2 023 1 channel per N/A A No N/A Reci rcul ation pump 1RSLOOPP152B Spray Flow 1RSLOOPP156A Outside 2 023 1 channel per N/A A No N/A Reci rcul ati on pump 1RSLOOPP156B Spray Flow NAPS TRM 3.3.9-14 Rev 75, 01/11/10
NUCLEAR DESIGN INFORMATION PORTAL Primary to Secondary Leakage Detection Systems B 3.4.5 B 3.4 REACTOR COOLANT SYSTEM B 3.4.5 Primary to Secondary Leakage Detection Systems BASES Primary to secondary leak monitoring is an important defense-in-depth measure to assure steam generator (SG) tube integrity. SG tube integrity is assured through the Reactor Coolant System (RCS) Operational Leakage and SG Tube Integrity Technical Specifications of 3.4, as well as the SG Program requirements of 5.5. NET 97-06 and its referenced EPRI Guidelines are the documents that define the SG Program referred to in the TS. Specifically, the PL1R Primary to Secondary Leak Guidelines (the Guidelines) define the program required by TS 5.5.8.e, A Steam Generator Program shall be established and implemented with provisions for monitoring operational primary to secondary LEAKAGE. Primary to secondary leak monitoring is one aspect of a programmatic, performance-based approach to ensuring SG tube integrity. A radiation monitor can be considered FUNCTIONAL if it is directly correlated to gpd leakage, can be monitored and will produce an alarm in the main control room, and can detect leak rates greater than 30 gpd at existing RCS activity levels. RCS activity levels existing during plant startup may not be adequate to support the required 30 gpd detection capability. One continuous radiation monitoring system consists of the N-16 continuous readout and alarm radiation monitors on each steam line. The second system consists of the N-16 continuous readout and alarm radiation monitor on the main steam header. The third system consists of the condenser air ejector exhaust continuous readout and alarm radiation monitor. The condenser air ejector exhaust monitor is correlated to gpd leakage by PCS or procedurally controlled hand calculation. Secondary system any secondary system sample that is mixed with the SGs during normal operation any individual SG sample, condensate system sample, feedwater system sample, main steam sample, or heater drain sample. Condition A, No Continuous Radiation Monitor FUNCTIONAL: A condition where there is no continuous radiation monitoring system sensitive to leak rates greater than 30 gpd and correlated to gpd for continuous on-line monitoring of primary to secondary leakage. This condition requires increased use of grab samples or other monitoring systems to ensure that a primary to secondary leakage event does not occur without rapid detection and response. Note that this condition will exist during plant startup until RCS activity increases to an adequate level to support the required 30 gpd detection capability. RCS N-16 activity will be sufficient to support FUNCTIONALITY of the N-16 systems with reactor power at or above 25%. RCS Ar-41 activity is expected to be sufficient to support FUNCTIONALITY of the condenser air ejector exhaust system shortly after 30% power is reached. A grab sampling program that is based on the t-ii dtL 44 I 5c) c
/
NAPS TRM t. if B 3.4.5-1 Rev 62,6K/O3/O7
NUCLEAR DESIGN INFORMATION PORTAL Primary to Secondary Leakage Detection Systems 3.4.5 3.4 REACTOR COOLANT SYSTEM 3.4.5 Primary to Secondary Leakage Detection Systems JR 3.4.5 The following primary to secondary leakage detection systems shall be FUNCTIONAL:
- a. One of the following continuous readout Radiation Monitoring Systems:
- The N-16 radiation monitoring system on each steam line, or
- The N-16 radiation monitoring system on the main steam header, or The condenser air ejector exhaust radiation monitor; and
- b. The capability to obtain and analyze one of the following:
- Condenser air ejector exhaust grab sample, or
- Liquid sample from the secondary system and from the Reactor Coolant System.
APPLICABILITY: MODES 1 and 2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Required continuous A.1 Obtain and analyze In accordance readout radiation grab samples. with monitor nonfunctional. Table 3.4.5-1 AND A.2 Restore required As soon as monitoring system to practical FUNCTIONAL. NAPS TRM 3.4.5-1 Rev 62, 05/03/07
NUCLEAR DESIGN INFORMATION PORTAL Regulatory Guide 1.97 Instrumentation 3.3.9 3.3 INSTRUMENTATION 3.3.9 Regulatory Guide (RG) 1.97 Instrumentation TR 3.3.9 RG 1.97 instrumentation for each function identified in Tables 3.3.9-1 (Unit 1) and 3.3.9-2 (Unit 2) shall be functional.
- NOTE - -
RG 1.97 requirements for Category 1 functions are stated in TS LCO 3.3.3. APPLICABILITY: Modes 1, 2, or 3.
- NOTES - -
RHR System Flow and RHR Heat Exchanger Outlet Temperature are also applicable when RCS is <350°F and <450 psig with fuel in the reactor vessel. Hi-Level Radioactive Liquid Tank Level, Radioactive Gas Hold-up Tank pressure and Radioactive Exposure Rate are applicable at all times. ACTIONS
- NOTE -
- 1. Separate Condition entry is allowed for each function identified in Tables 3.3.91 (Unit 1) and 3.3.9-2 (Unit 2).
- 2. Table 3.3.9-3 lists alternate indication for Category 2 and 3 Functions.
NAPS TRM 3.3.9-1 Rev 75, 01/11/10
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 42. 056AG2.4.9 042/NEW//H/3/3 .8/4.2/6/
Given the following conditions:
- Unit 1 RCS cooldown is in progress due to leakage from a PRZR manway.
- A RHR pump is in service.
- RCS temperature is 190°F with a 30°F/Hr cooldown rate.
A loss of F transfer bus occurs. In accordance with 1-AP-1 1, Loss of RHR, in order to re-establish RHR cooling after the EDG energizes the bus, the crew is directed to A. verify A RHR pump running, and manually adjust 1-CC-MOV-100A & B, CC HX Outlet Isolation Valves. B. verify B RHR pump automatically starts, and manually adjust 1-RH-HCV-1 758, RHR HX Outlet Valve. C. manually start A RHR pump, and manually adjust 1-RH-HCV-1 758, RHR HX Outlet Valve. D. manually start A RHR pump, and manually adjust 1-CC-MOV-100A & B, CC HX Outlet Isolation Valves.
- a. Incorrect. Plausible since the candidate who lacks detailed system knowledge may confuse operation of the RHR pump with that of the LHSI pump, or other pumps (i.e. Charging pump) that ride the bus.
The candidate also may not understand that CC pumps have an auto start feature, and while the given action is a subsequent action it would be appropriate only if there were issues with the CC system operation (not the case based on the information provided).
- b. Incorrect. Plausible since the candidate who lacks detailed system knowledge may confuse operation of the RHR pump with that of the CC system pumps, or other pumps that have auto starts from loss of their associated running pump. Second part is correct.
- c. Correct. As discussed above if all systems work as designed the operator will have to start the subject pump. Second part is also correct, but the candidate who lacks knowledge of the substeps for restoring flow may consider this action unnecessary and thus conclude it to be incorrect.
- d. Incorrect. First part is correct as explained above. Second part is incorrect but plausible as discussed in distractor a.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Loss of Offsite Power Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies. (CFR: 41.10/43.5 /45.13) Tier: 1 Group: 1 Technical
Reference:
1-AP-li Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
NUMBER ATTACHMENT TITLE ATTACHMENT 0-AP-lO 24 UNIT 1 CR0 LOSS OF POWER ACTIONS REVISION PAGE 62 2ofll _STEP ACTION/EXPECTEDRESPONSE I I RESPONSENOTOBTAINED I CAUTION: If CC is unavailable to operating RCPs, then RCP temperatures should be closely monitored. 2._ CHECK CORE COOLING AVAILABLE:
- H RHR should be in service, THEN initiate 1-AP-il, LOSS OF RHR.
D
- At least one RCP - RUNNING D IF Unit 1 OR Unit 2 PCS is available, THEN monitor the natural circulation parameters.
D
- AtleastoneRHRPump-RUNNING IF the Unit 1 PCS Unit and 2 PCS are NOT available, THEN monitor the following:
D
- RCS subcooling based on Core Exit TCs are greater than 25 °F
- SG pressures are stable or decreasing D
- RCS Hot Leg temperatures are stable or decreasing
- Core Exit TCs are stable or decreasing D
- RCS Cold Leg temperatures are at the saturation temperature for SG pressure
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Donanen NORTH ANNA POWER STATION ABNORMAL PROCEDURE NUMBER PROCEDURE TITLE REVISION 27 1-AP-li LOSSOFRHR (WITH ELEVEN ATTACHMENTS) PAGE 1 of 23 PURPOSE To provide instructions for maintaining Core Cooling and protecting the Reactor Core in the event that RHR Cooling is lost. ENTRY CONDITIONS This procedure is entered when RHR is required for Core Cooling and any of the following conditions exist:
- Air-binding of operating RHR pumps as indicated by:
- Flow oscillations, or
- Motor amps fluctuating, or
- Excessive pump noise.
- Annunciator E Panel A-6, RHR PP 1A AUTO TRIP, is LIT, or
- Annunciator E Panel A-7, RHR PP 1 B AUTO TRIP, is LIT, or
- Annunciator E Panel A-8, RHR SYSTEM LO FLOW, is LIT, or
- Loss of RHR pumps due to loss of power, or
- Failure of RHR system to control RCS temperature due to loss of CC or valve failures, or
- Loss of Service Water System with RHR System in service, or
- Loss of Component Cooling System with RHR System in service.
CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 27 1-AP-il LOSSOFRHR PAGE 2of23 -_STEP ACTION/EXPECTED RESPONSE j RESPONSE NOT OBTAINED CAUTION:
- RCS make-up concentration MUST be greater than or equal to current Shutdown Margin and Boron concentration requirements of the COLR.
Changes in RCS pressure due to boiling in the core can result in Reactor Vessel water level changes that may not show on RCS standpipe level indicator 1 -RC-LI-1 03. CHECK RCS LEVEL - DECREASING C GO TO Step 5. C
- RCS standpipe level -
DECREASING OR C
- RCS ultrasonic level indicator -
DECREASING OR C
- PRZR level - DECREASING OR C
- RCS makeup rate -
INCREASING OR C
- Containment Sump pumping frequency UNEXPLAINED INCREASE OR C
- PDTT pumping frequency -
UNEXPLAINED INCREASE 2._ INCREASE RCS MAKEUP FLOW
NUMBER PROCEDURE TITLE REVISION 27 1 -AP-1 1 LOSS OF RHR PAGE 3 of 23 -_STEP ACTION/EXPECTEDRESPONSE I RESPONSENOTOBTAINED 3._ ISOLATE RCS DRAIN PATHS: a) Check the following Letdown C a) Manually close valves. Isolation Valves CLOSED: C
- 1-CH-HCV-1200A C
- 1-CH-HCV-1200B C
- 1-CH-HCV-1200C C
- 1-CH-LCV-1460A C
- 1-CH-LCV-1460B C b) Check 1 -CH-HCV-1 142, RHR C b) Manually close valve.
System to Letdown Isolation Valve
- CLOSED c) Check loop drains - CLOSED: C c) Manually close valves.
C
- 1-RC-HCV-1557A C
- 1-RC-HCV-1557B C
- 1-RC-HCV-1557C d) While continuing with procedure, C d) Ensure valves are closed.
verify the following valves - LOCKED CLOSED: C
- 1-RH-36, Residual Heat Removal to RWST Isolation Valve (Containment)
C
- 1-RH-34, Residual Heat Removal Supply to RP (Containment)
(STEP 3 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 27 1-AP-li LOSSOFRHR PAGE 4of23 H_STEP ACTION/EXPECTED RESPONSE j j RESPONSE NOT OBTAINED
- 3. ISOLATE RCS DRAIN PATHS:
(Continued) e) Close any known RCS drain paths EJ f) Initiate actions to stop level decreases due to maintenance covered by 0-GOP-i 3.3, ASSESSMENT OF MAINTENANCE ACTIVITIES FOR POTENTIAL LOSS OF REACTOR COOLANT INVENTORY CAUTION:
- RHR flow less than the design flow indicated by ATTACHMENT 3 may cause RCS temperature to increase.
Changes in RCS pressure can result in Reactor Vessel water level changes that may not show on RCS level indicator 1-RC-LI-i03.
- 4. VERIFY ADEQUATE RCS MAKEUP FLOW:
D a) Check RCS level - STABLE OR a) Ensure the keylock switch for i-RC-LI-i05, INCREASING Independent RCS Level Indicator, is in ENABLE. GO TO appropriate procedure: D
- 1-AP-i7, SHUTDOWN LOCA OR D
- i-AP-52, LOSS OF REFUELING CAVITY LEVEL DURING REFUELING (STEP 4 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 27 1-AP-li LOSSOFRHR PAGE 5 of 23 -_STEP ACTION/EXPECTED RESPONSE J I RESPONSE NOTOBTAINED
- 4. VERIFY ADEQUATE RCS MAKEUP FLOW: (Continued) b) Check RHR flow LESS THAN OR
- D b) Reduce RHR flow to design flow rate of EQUAL TO DESIGN FLOW OF ATTACHMENT 3.
ATTACHMENT 3:
- 2RHRHXsin use-Page 1 of 2 D
- lRHRHXinuse Page 2 of 2 D c) Check RCS level GREATER
- c) Do the following:
THAN MINIMUM FOR INDICATED FLOW OF ATTACHMENT 2 D 1) Continue RCS makeup. EJ 2) Stop RHR Pumps. D 3) GOTOStep11. d) Check RCS level AT LEAST
- d) Increase RCS level to greater than +10 inches
+10 INCHES ABOVE above centerline.
CENTERLINE IF level cannot be increased to greater than
+10 inches above centerline, THEN GO TO 1-AP-17, SHUTDOWN LOCA.
NUMBER PROCEDURE TITLE REVISION 27 1-AP-il LOSSOFRHR PAGE 6 of 23 _STEP ACTION/EXPECTEDRESPONSE j RESPONSENOTOBTAINED
- 5. VERIFY RHR ISOLATION VALVES -
OPEN: a) RHR Inlet Isolation Valves - OPEN a) Do the following:
- 1-RH-MOV-1700 D 1) Stop RHR Pump(s).
D
- 1-RH-MOV-1701 I 2) Reduce RCS pressure as necessary.
D 3) WHEN RCS pressure is less than 418 psig, THEN open valves. b) AT least one RHR Outlet Isolation lJ b) Open at least one RHR Outlet Isolation Valve. Valve OPEN C
- 1-RH-MOV-1720A C
- 1-RH-MOV-1720B CAUTION: RHR flow less than minimum requirements may cause RCS temperature to increase.
NOTE:
- Operating at low RHR system flow rates during reduced inventory operations greatly reduces the risk of air entrainment (vortexing).
- Indications of a pump sheared shaft are low flow and low motor amps. A degraded pump or a pump with a sheared shaft is to be considered as NOT running.
CHECK ONE RHR PUMP - RUNNING Do the following: C a) IF the other RHR pump is available, THEN stop any degraded RHR pump. (STEP 6 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 27 1-AP-il LOSSOFRHR PAGE 7 of 23 ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED
- 6. CHECK ONE RHR PUMP -
RUNNING (Continued) D b) IF a degraded RHR pump is running AND the other RHR pump is NOT available, THEN GOTOStep7. c) IF electrical power is available, THEN do the following:
- 1) Manually close the following RHR Control Valves:
D
- 1-RH-FCV-1605
- 1-RH-HCV-1758 D 2) IF an RHR Pump was previously stopped due to air entrainment, THEN locally vent both RHR Pumps.
D 3) IF both RHR pumps are stopped, THEN start one RHR pump.
- 4) Restore RHR flow by repositioning the following RHR Control Valves:
D
- 1-RH-HCV-1758 D
- 1-RH-FCV-1605 D 5) IF an RHR Pump has been started, THEN GOTO Step 7.
no RHR Pump can be started, THEN GOTO Step 11. (STEP 6 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 27 1-AP-il LOSSOFRHR PAGE 8 of 23 -_STEP FI ACTION/EXPECTEDRESPONSE I RESPONSENOTOBTAINED
- 6. CHECK ONE RHR PUMP -
RUNNING (Continued) d) IF electrical power is NOT available, THEN do the following:
- 1) Initiate 0-AP-lO, LOSS OF ELECTRICAL POWER.
D 2) GOTOStep11.
- 7. VERIFY RHR SYSTEM - NORMAL: Do the following:
D
- RHR flow - NORMAL a) IF RHR Pump is vortexing, THEN do the following:
D
- RHR flow - STABLE D 1) Start increasing RCS level to at least D
- RHR Motor amps - STABLE +10 inches above centerline by increasing charging flow.
- RCS temperature - STABLE
- 2) Check RHR flow less than or equal to design flow of ATTACHMENT 3:
D
- 2RHRHXsinuse-Pagel of2 D
- 1 RHRHXinuse-Page2of2 IF RHR flow is greater than the design flow rate of ATTACHMENT 3, THEN reduce flow to the design flowrate using:
- 1-RH-HCV-1758 D
- 1-RH-FCV-1605 (STEP 7 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 27 1-AP-il LOSSOFRHR PAGE 9of 23 H_STEP ACTION/EXPECTEDRESPONSE I I RESPONSENOTOBTAINED
- 7. VERIFY RHR SYSTEM - NORMAL:
(Continued) 1 3) Check RCS level Greater than minimum for indicated flow of ATTACHMENT 2. D IF RCS level is not greater than minimum for indicated flow of ATTACHMENT 2, THEN STOP the RHR Pumps and GO TO Step 11.
- 4) Send an Operator to locally check pump operation:
D
- RHR pump noise LJ
- RHR pump seals D
- RHR pump vibration D b) IF the running RHR pump is degraded AND the other RHR pump is available, THEN RETURN TO Step 6.
D c) IF RHR System cannot be stabilized, THEN stop running RHR Pump AND GO TO Step 11.
NUMBER PROCEDURE TITLE REVISION 27 1-AP-li LOSSOFRHR PAGE 10 of 23 -j_STEP j ACTION/EXPECTEDRESPONSE I I RESPONSENOTOBTAINED I
- 8. CHECK SERVICE WATER TO CC HEAT EXCHANGER AVAILABLE:
D a) Verify Service Water System - IN a) IF Service Water flow is NOT available, THEN SERVICE initiate the following while continuing with this procedure: D
- 0-AP-12, LOSS OF SERVICE WATER D
- 1-AP-15, LOSS OF COMPONENT COOLING E1 GOTOStep11.
b) Verify Service Water Supply Valves b) Open Service Water Supply Valves to CC to CC System OPEN:
- System:
LJ
- 1-SW-MOV-108A D
- 1-SW-MOV-108A D
- 1-SW-MOV-108B D
- 1-SW-MOV-108B EJ c) Locally check Service Water to CC D c) GO TO Step 11.
Heat Exchanger P NORMAL
NUMBER PROCEDURE TITLE REVISION 27 1-AP-il LOSSOFRHR PAGE ii of 23 ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED 9._ CHECK CC FLOW TO RHR HEAT Do the following: EXCHANGERS NORMAL: D
- 1-CC-FI-132A a) Open CC valves for in service CC Heat Exchanger:
D
- 1-CC-FI-132B
- i-CC-TV-iO3A, A RHR Heat Exchanger Return Isolation
- 1-CC-TV-103B, B RHR Heat Exchanger Return Isolation
- 1-CC-MOV-100A, A CC Heat Exchanger Outlet Isolation D
- i-CC-MOV-100B, B CC Heat Exchanger Outlet Isolation b) IF either 1-CC-TV-i 03A or 1-CC-TV-i 03B cannot be opened, THEN close the associated RHR CC MOV:
D
- 1 -CC-MOV-i OOA for i-CC-TV-i 03A
- i-CC-MOV-100B for i-CC-TV-i 03B D c) HZ CC flow is restored, THEN GO TO Step 10.
1 d) IF CC is NOT restored, THEN initiate i-AP-i5, LOSS OF COMPONENT COOLING, while continuing with this procedure. e) GOTOStepii.
NUMBER PROCEDURE TITLE REVISION 27 1-AP-il LOSSOFRHR PAGE 12 of 23 H_STEP [j ACTION/EXPECTEDRESPONSE I I RESPONSENOTOBTAINED I i0. RETURN TO PROCEDURE AND STEP IN EFFECT CAUTION: If RCS boiling is determined to exist, then non-essential personnel should be evacuated from the Containment. i1._ INITIATE PERSONNEL PROTECTIVE ACTIONS: D a) Record most recent time to boiling estimate from 1-GOP-i 3.0, ALTERNATE CORE COOLING METHOD ASSESSMENT: Time (minutes): b) Evaluate need to implement EPIP-i .Oi, EMERGENCY MANAGER CONTROLLING PROCEDURE c) Monitor Containment Radiation:
- 1-RM-RMS-159
- 1-RM-RMS-160 i2. INITIATE ATTACHMENT ii, CONTAINMENT CLOSURE, WHILE CONTINUING WITH THIS PROCEDURE
- 13. VERIFY 1 -RC-LI-i 05, INDEPENDENT D Place the keylock switch for 1 -RC-LI-1 05 in RCS LEVEL INDICATOR - ENABLE.
ENERGIZED
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 43. 057AG2.4.45 043/NEW//H/4/4. 1/4.3/6/
Unit 1 is at 100% power when several Unit 1 annunciators are received. The OATC verifies all critical parameters are stable and on program. The crew notes the following:
- All status lights on L Panel are OFF.
- Annunciator 1P-E4, C-7 PERM STM DUMP ARMED FROM LOSS OF LOAD, is LIT.
Which ONE of the following identifies the 12OVAC Vital bus that has lost power, AND includes the action needed to reset (clear) 1 P-E4 once power is restored? A. Vital bus I-Ill; momentarily place the Steam Dump Mode Selector in the RESET position. B. Vital bus I-Ill; place both Steam Dump INTLK switches to OFF/RESET and return to ON. C. Vital bus -IV; momentarily place the Steam Dump Mode Selector in the RESET position. D. Vital bus -IV; place both Steam Dump INTLK switches to OFF/RESET and return to ON.
- a. Incorrect. Plausible since this vital bus interfaces with both the instrumentation and arming circuit for the steam dumps and the candidate who lacks a thorough understanding of this and the annunciator system power supplies may default to this distractor. Second part is correct, as noted in the annunciator response this feature has a retentive memory that can only be reset with this switch.
- b. Incorrect. First part incorrect but plausible as noted above. Second part is incorrect but plausible especially since the switch position (OFF/REST) implies that it would perform the subject function.
- c. Correct. First part is correct, loss of the bus will effect these status lights because of loss of the demultiplexer and although a loss of load has not occurred the failure will result in an arming signal.
Second part is correct as discussed in distractor a.
- d. Correct. First part is correct as discussed in answer c. Second part is incorrect but plausible as discussed in distractor b.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Loss of Vital AC Electrical Instrument Bus Ability to prioritize and interpret the significance of each annunciator or alarm. (CFR: 41.10 / 43.5 / 45.3/45.12) Tier: 1 Group: 1 Technical
Reference:
AR P-E4 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
VIRGINIA POWER 1-EI-CB-21P ANNUNCIATOR E4 1-AR-P-E4 NORTH ANNA POWER STATION REV. 1 APPROVAL: ON FILE Effective Date:07/30/03 C7 PERM Turbine load decreasing STM DUMP by greater than 10 ARMED FROM percent in 120 seconds LOSS OF LOAD NOTE: The Steam Dump Armed from Loss of Load permissive is cleared by momentarily placing the Steam Dump Mode Selector in the RESET position. 1.0 PROBABLE CAUSE Turbine load rejection
2.0 REFERENCES
2.1 NADW5655D33, Nuclear Steam Supply System, Functional Diagrams, Steam Dump Control, Units 1 and 2, Sheet 10 of 16 2.2 1l715ESK1OCAF, TripPermissive and Bypass Status Lights, Sheet 6 of 8 2.3 117l5FE-7X, Wiring Diagram Annunciator, Demultiplexer ANJ64 to ANJ67 2.4 11715MS131, Main Steam System, Turbine First Stage Pressure Channel IV 2.5 11715MS172, Main Steam System, Low Turbine 1st Stage Pressure Channel IV 2.6 ICPP1--P447, P447 First Stage Pressure Protection Channel IV 2.7 NCRODP-23, Main Steam System 2.8 NCRODP-75, Main Turbine Generator Control and Protection System 2.9 NCRODP-77, Reactor Protection System 3.0 ACTUATION lMSPC1447AX gtW (7 ,
/ - --V y---- L I C
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 44. 05 9A 1.07 044/NEW//L/4/2.7/2 .9/4/
1-OP-2.1, Unit Startup from Mode 2 to Mode 1, is in progress. Operators are establishing conditions to synchronize and load the Main Generator. Prior to synchronizing and loading the Main Generator, 1-OP-2.1 requires Condensate Pump(s) running, and FRV Bypass valves A. ONLY one ; in AUTO with SG levels on program. B. ONLY one; in MANUAL with SG levels at approximately 40%. C. two; in AUTO with SG levels on program. D. two; in MANUAL with SG levels at approximately 40%.
- a. Incorrect. First part is plausible since this is all that is physically required (the second pump is desired to be running to provide additional margin for FW pump suction pressure). Second part is also plausible since for most conditions auto is preferred, this situation is unique in that manual control is preferred in order to establish additional margin from the SG low-level trips.
- b. Incorrect. First part incorrect but plausible as noted above. Second part is correct this is done to place SG levels closer to the middle of the band thus providing additional margin from the SG low-level trips.
- c. Incorrect. First part is correct 1-OP-2.1 does this to minimize the potential of an auto start that could result in a significant secondary swing. Second part incorrect but plausible as discussed in Distractor a.
- d. Correct. First part is correct as discussed in Distractor c. Second part is also correct as discussed in distractor b.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Main Feedwater (MFW) System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MEW controls including: (CFR: 41.5 / 45.5) Power level restrictions for operation of MFW pumps and valves. Tier: 2 Group: 1 Technical
Reference:
1-OP-2.1 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info: considered to meet K/A intent since question goes beyond level of detail required by K/A and tests procedural knowledge of how MEW controls (FRV Bypass valves) are operated for specific plant conditions. Additionally all three MEW pumps at NAPS are essentially indentical in both construct and capacity, thus there is no preferred pump for evolutions such as Unit startup and the like.
PROCEDURE NO: r Wp Daminiet 1-OP-2.1 REVISION NO: NORTH ANNA POWER STA TION 95 PROCEDURE TYPE: UNIT NO: OPERATING PROCEDURE PROCEDURE TITLE: UNIT STARTUP FROM MODE 2 TO MODE I SURV ICCE REACT REQ III MGT REVISION
SUMMARY
Made the following changes for DCP 07-02 1, Feedwater UFM Installation PCS - Revised synopsis statement calorimetric power (U1203) to calorimetric power (U1231, Power, 1 Mm Avg UFM or U1203, Power, 1 Mm Avg FW). Added a table to the synopsis to show power level indications. Added U1231, Power, 1 Mm Avg UFM to P&L 4.25, Caution for Step 5.2.1 and Caution for Step 5.2.79. In Step 5.2.9(c) Changed Operations Manager on Call to Senior Operations Manager because OMOC is not required for ICCE III per OP-AA-106. In Step 5.2.56 added cross reference to Step 5.2.55 to identifi which annunciators. PROBLEMS ENCOUNTERED: jj NO YES Note: If YES, note problems in remarks. REMARKS: (Use back for additional remarks.) SRO: DATE: CONTINUOUS USE
DOMINION 1-OP-2.1 North Anna Power Station Revision 95 Page 31 of98 NOTE: The measurement of Air Ejector flow rate using the Rotameters has potential inaccuracies at high Air Ejector flow rates. The Rotameter measurements are effective in identifying higher than desired Air Ejector flow rates, but the numerical value of this flow rate may not be as accurate as the lower flow rates typically observed. (Reference 2.3.57) 5.2.15 Check the Air Ejectors, as follows: (Reference 2.3.57)
- a. Measure Air Ejector flow rates using l-OP-36.2, Main Condenser Air Ejectors and record below:
- 1-CN-EJ-1A Flow Rate: SCFM
- 1-CN-EJ-1B Flow Rate: SCFM
- Total Air Ejector Flow Rate: SCFM
- b. IF total Air Ejector flow rate is greater than 6 scfrn, THEN have the OMOC determine if the source of inleakage should be determined prior to power escalation.
5.2.16 WHEN the 5% Chemistry hold has been cleared, THEN continue with this procedure. 5.2.17 Do the following to prepare for the anticipated increase in Feedwater and Condensate flow: I
- a. Start a second Condensate Pump using 1-OP-30.1, Operation of Condensate System.
NOTE: 1-FW-FCV-150C, C Main Feedwater Recirc Header Flow Control Valve, is the preferred recirc to remain open, as it will provide more efficient Hotwell recirculation, dumping into the A Condenser on the East end, furthest away from the Condensate Pump suction on the B Condenser.
- b. Initiate Attachment 5, Control of Feed Water Recircs During Power Ascension.
DOMINION 1-OP-2.1 Noh Anna Power Station Revision 95 age o 7 NOTE: The Bypass Feedwater Regulating Valves must be in Manual to establish
/ approximately 40 percent SG level. In Auto, the Bypass Feedwater
( Regulating Valves will attempt to maintain approximately 33 percent SG level due to the low First Stage Pressure signal.
- i. Establish the Narrow Range SG level at approximately 40 percent and match Steam Flow and Feed Flow with the Bypass Feedwater Regulating Valves.
- j. Verify Tave is stable.
5.2.24 Verify Step 5.2.10 has been completed. CAUTION IF MFW temperatures are excessively low, THEN severe SG level oscillations may occur when the Main is placed in service. Generator 5.2.25 Verify the following: Step 5.2.23 has been completed. Conditions are stable. 5.2.26 IF two Main Condensate Pumps are NOT running, THEN start a second Main Condensate Pump using 1-OP-30.1, Operation Of The Condensate System. c
DOMINION 1-OP-2.1 North Anna Power Station Revision 95 Page 36 of 98 CAUTION To prevent anti-motoring trip, the ramp rate should be 2% per minute until positive indication of load increase is observed on the megawatt recorder. (References 2.4.8 and 2.4.11) 5.2.27 Adjust the setter ramp rate to 2% / minute. (References 2.4.8 and 2.4.11) chzd load the ain Generator using 1 -OP-15.2, Main Generat:? 5.2.29 WHEN positive indication of load increase is observed on the megawatt recorder, THEN reduce the setter ramp rate to 0.3% per minute. NOTE: Control Rod Insertion Limits are a function of Median/High Control AT. 5.2.30 Operate the blender and Control Rods in manual as required to maintain the following: Tave and Tref within 1.5 degrees of each other.
- Control Rods above the insertion limits.
5.2.31 Give Attachment 8, Guidance For Main Turbine Operations, to the RO operating the Main Turbine. 5.2.32 Start increasing Main Turbine power, while doing the following:
- Monitor NIs.
- Monitor Core Average AT.
- Monitor Tave and Tref.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 45. 059K1.04 045/NEWI/H/3/3.4/3.4/4/
Given the following conditions:
- Unit 1 is at 25% power and ramping up.
- Unit 2 is in a refueling outage.
- All Unit 1 SGWLC channels are selected to Channel Ill (Blue channel).
An I&C Tech inadvertently equalizes and isolates 1-FW-FT-1487, Unit 1 B SG Feed Flow Transmitter (Blue channel). Which ONE of the following describes the result of the I&C technicians error? (assume NO operator action is taken) A. B SG level decreases until the turbine automatically trips. B. B SG level decreases until the reactor automatically trips. C. B SG level increases until the turbine automatically trips. D. B SG level increases until the reactor automatically trips.
- a. Incorrect. Plausible since the candidate may confuse (equate) the response as similar to that of a SG level transmitter with a reference leg leak; the response is however opposite. Also the candidate who lacks detailed knowledge of the RPS system may conclude that this trip is a P-8 interlocked trip which is not the case.
- b. Incorrect. Plausible as noted above and if the failure mode were correct an automatic Reactor trip would result. Also, reactor trip from turbine trip is blocked below P-8.
- c. Correct. The failure mode results in FW flow of zero so the FRV will open to compensate and at 75%
level an automatic turbine trip on hi-hi SG level will result; this signal also generates a FW isolation signal which is why SG level increase is terminated.
- d. Incorrect. Plausible because a reactor trip is an eventual result due to the loss of all feed discussed in answer c; again this is a case where the candidate who lacks detailed knowledge may erroneously conclude that either a SG hi-hi level or all MFW pump breakers open actuates Rx trip logic, however this is not the case. Also, reactor trip from turbine trip is blocked below P-8.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Main Feedwater (MFW) System Knowledge of the physical connections and/or cause effect relationships between the MFW and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8) S/GS water level control system Tier: 2 Group: 1 Technical
Reference:
5655D33 sh. 13 of 16 and sh. 7 of 16 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
STEAM GENERATOR iI STEAM GENERATOR STEAM GENERATOR i3 PROGRAMMED PROGRAMMED PROGRAMMED LEJEL SETPDINT c LEVE SETPI\ LEVEL S TPOINT STEAM STEAM FEEDWATER FEEDWATEP AM STEAM FEEDSATEr, EEDWATER STEAM STEAM FEEDWATER FEEDWATER FLOW FLOW FLOW FLOW FLOW FLOW FLOW FLOW FLOW FLOW FLOW FLOW
- (!_, (L ÷
+ (D I I I I I+t V S I+t t - )+t H-t 32 C H II I*
I I
) C> D C Q
-: POWER -: -:
RANGE REDUNDANT I I I
- I- I I K 3 Ir I K 4) 36 t
S K 36 t SS 3 i K 1 1 I (1÷ K 31 (1+ 1 K 31 (1÷
- 33 I
-- 1 - --
AUTO-MANUAL AU 10-MANUAL WIDE RANGE LEVEL CHANNELS AUTO-MANUAL Au 0-MANUAL AUTO-MANUAL AUTO-MANUAL CONROL CONTROL CONTROL CONTROL CONTROL CONTROl 5A HON STATION STEAM STEAM STEAM S TA (ION STATION STATION STATION GEN I SEN 2 GEN 3 TRIP ALL FFFDWA TER PUMPS TURBINE TRIP SHEET 151 ISLOATION VALVES RN
*--()
LVE o--- t NOTES 3 5 5) MOV-FWI5RA.B & C , .. AUX AUX AUX STAHON STATiON STATION
/EEDATER (A PUMP DISCHARGE I REDUNDANT RE DUNDA N F REDUNDANT M01 N TRAIN B ONLY (NOTE II (NOTE I) (NO I E I) (NOTE N
NOTES: REDUNDANT ANALOG GATE CONSISTS OF 2 SOLENOID VENT \ALVES IN SERIES TO MODUL ATE MODULATE MODULATE MODULA TB MODUL ATE MODULATE RECUiDANTLY INTERLOCK THE AIR LINE BETWEEN EACH VALVE FEEDWATER FEEDWATER FECUWA TSR FEED (V A TSR FEEDWATER FEED WATER DIAPHRAGM AND ITS ASSOCIATED POSITIONER. THE SOLENOID VALVES MAIN VALVE BYPASS VALVE MAIN VALVE BYPASS VALVE MAIN VALVE BYPASS VALVE ARE 3D-ENERGIZE TO VENT, CAUSING THE FEEDWATER VALVE TO CLOSE ECV-178 BY OTHERS FL V -188 (BY OTHERS) FCV-1V8 (BY OTHERS) IN 5 SECONDS. (NOTE 6) 501-F WI55 A (NOTE 6) SOy -FW[55B (NOTE 61 SOV FWI55C (NOTE 61 (NOTE SI (NOTE 6)
- 2. ALL CIRCUITS UN THIS SlEET ARE NOT REDUNDANT, EXCEPT WHERE INDICATED REDUNDANT.
- 3. TRIPPING CF FEEDWATER PUMPS CAUSES CLOSURE OF ASSOCIATED PUMP DISCHARGE VALVES.
FF NUCLEAR STEAM SUPPLY SYSTEM
- 4. THE REDUNDANT MANUAL RESET CONSISTS OF TWO MOMENTARY CONTROLS ON THE CONTROL BOARD, ONE FOR EACH TRAIN. C fo 0 LC)
Cr, FUNCT1ONL DIGRMS FEE0WTER CONTROL & ISOLT ION S. THE FEEDWATER 0 BY OTHERS. UMPS AND PUMP DISCHARGE VALVES ARE SUPPLIED
- 8. OPEN/SHIJT IN4DICA (ION FOR EACH FEEDWATER VALVE (N CONTROL ROOM.
F
\ Oi.i cQ-,
1 A)e L%e 0) Cr, 10 10 10 10 IJNITS 1 & 2 c IL)
.A&ç 1 \_ 10
)O K) f-v () ev NUCLEAR ENGINEERING Cl Dominion NORTH ANNA POWER STATION
- PEVISEO PER OCR 2007-0857 O\IM I Z LI)
THIS DWO SUPERSEDES REV 0 o LI)
.0 AEVISION EASCRIPTION ASCiI ci LI)
C) Z N- OW- 5655033 13161 1 PCNO1 05-FEB-200W lSø2 PRIOR TO USING FOR DESIGN WORK CHECK OMIS FOR WORK PENDING
HIGH STEAM LINE DIFFERENTIAL PRESSURE HI STEAM LINE FLOW LOW STEAM LINE PRESS. P1-P2 PLP2 P1-P2 P2-P3 P2-P3 P2-P3 Pl-P3 Pi-PO P1-P3 LOOP 1 LOOP 2 LOOP 3 [OP 1 LOOP 2 LOOP 3 LOOP I II II III III IV IV II II III III IV IV II II III III IV IV III IV III IV III IV II III IV P 12 (1E) LDLO T AVG HOT LEG STOP VALVE COLD LEG STOP VALVE () C 000 PC PC LOOP 2 00 000 PC PC PC 474 FE 475 PC 485 FC 494 Ft 495 pc 485 PC 496 (SHEET 5) CLOSED 33BC CLOSED 33BC HOT LEG CDLDLEG HOT LEG COLD LEG JhL i 0 J VALVE STOP VALVE STOP VALVE STOP VALVE
Ø ._Ø 0 Q STOPCLOSED CLOSED CLOSED 33BC CLOSEC 33BC T T KD T T T fJ c 33BC A 2 I A 2 fA A 2 A 33BC TE 1) 23 a-A, 2 ( NOTE 1 ( A3 -A, -A, c c c __)iJ __( T
-HL --1-1 HL --HL c-Ø A 8 6
A A I( STEAMLINE LINE SI MANUAL SI MANUAL 2,/ RESET BLOCK (NOTE 4) (NOTE 2) SET BLOCK jE_ 2/ I II MOMENTARY) A P3 (NOTE 3) TO STEAM LINE C ISOLATION (SHEET 8) 4k C P OTHER TO SAFETY R LOGIC INJECTION TO SAFETY TRAIN (SHEET 8) INJECTION LOW FEEDWATER FLOW (SHEET 8) STEAM SEN. LOW-LOW WATER LEVEL STEAM SEN. HI-HI LEVEL LOOP 1 LOOP 2 LOOP 3 LOOP 1 LOOP 2 LOOP 3 STM.GEN. 5TM.GEN. 2 STM.GEN. 3 I II III I II III I II III II II LC 474 LU 475 LC 476 LU 484 LU 485 LC 486
/EN (494)
/EN 1495)
LU (498) 6 4 8 LL L T T T C C C S S S 10 10 13 2/
/3 NOTES:
- 1. OSITION DETECTION IS ACCOMPLISHED BY 2 SWITCHES (INDIVIDUAL FOR EACH TRA:N( PER STOP VALVE.
- 2. THE REDUNDANT MANUAL BLOCK CONTROL CONSISTS OF TWO CONTROLS ON THE CONTROL BCARD, ONE FOR EACH TRAIN.
- 3. TWO COMPUTER INPUTS ARE CONNECTED TO THIS CIRCUIT.
INDIVIDUAL FOR EACH TRAIN. NUCLEAR STEAM SUPPLY SYSTEM
- 4. THE REDUNDANT MANUAL RESET CONTROL CONSISTS OF TWO FUNCTIONAL DIAGRAMS CONTROLS ON THE CONTROL BOARD, ONE FOR EACH TRAIN. Vi STEAM GENERATOR TRIP SIGNALS Vi TO AUXILIARY P- 14 TO Vi FEEDWATER PUMP TURBINE TRIP & Vi START-UP LOGIC FEEDWATER ISOLATION S-I (SHEET 141 (SHEET 13) Vi ST EQUIPMENT CLASSIFICATION ON THIS DWG. IS SAFETY RELATED C Vi Cr)
CE) REACTOR TRIP o (SHEET 2) zz z VIRGINIA POWER
-, C 1 REVISED PER DC 88-01 I I I I I 0 I OR JO I NAL I SSUE JAR I 8- l 30) 00 NORTH ANNA POWER STATION i 1 CMF THIS DWG. SUPERSEDES REV 0 VRIGINVL I LL JJG 86 REV DESCRIPTION DOWN CARD DSPL DATE REV DESCRIPTION [9RWN CHKO DSPL lATE PCNOI NA- OW- 5655D33 3H7 OF 16 - I I 7-OCT- 1997 15:52 PRIOR TO USING FOR DESIGN WORK CHECK DMIS FOR WORK PENDING
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 46. 061 G2. 1.23 046/BANK/NAPS: 7065/L/3/4.3/4.4/4/
Given the following conditions:
- Unit 1 experiences a reactor trip and safety injection.
- 1-FW-P-2, Turbine Driven AFW pump, will not operate.
The crew performs 1-AP-22.1, Loss of 1-FW-P-2 Turbine Driven AFW Pump, to feed all three SGs. Which ONE of the following identifies the AFW alignment once 1-AP-22.1 has been performed? A. 1-FW-P-3A and 1-FW-P-3B feeding A SG thru 1-FW-HCV-100A 1-FW-P-3A and 1-FW-P-3B feeding B SG thru 1-FW-HCV-100B 1-FW-P-3A and 1-FW-P-3B feeding C SG thru 1-FW-HCV-100C B. 1-FW-P-3A feeding A SG thru 1-FW-HCV-100A 1-FW-P-3B feeding B SG thru 1-FW-MOV-100B 1-FW-P-3A feeding C SG thru 1-FW-HCV-100C C. 1-FW-P-3B feeding A SG thru 1-FW-MOV-100A 1-FW-P-3B feeding B SG thru 1-FW-MOV-100B 1 -FW-P-3A feeding C SG thru 1 -FW-HCV-1 OOC D. 1-FW-P-3A feeding A SG thru 1-FW-MOV-100D 1-FW-P-3B feeding B SG thru 1-FW-MOV-100B 1-FW-P-3A feeding C SG thru 1-FW-HCV-100C
- a. Correct. For this condition an abnormal alignment is established as described with both pumps and all SGs aligned to the HCV header.
- b. Incorrect. Plausible since this alignment would remedy the situation and maintain a degree of train separation, however this is not lAW the subject procedure.
- c. Incorrect. Plausible as discussed in Distractor b.
- d. Incorrect. Plausible as discussed in Distractor b.
QUESTIONS REPORT for NAPS 2010 RD NRC Exam rev3 Auxiliary / Emergency Feedwater (AFW) System Ability to perform specific system and integrated plant procedures during all modes of plant operation. (CFR: 41.10 /43.5 /45.2 /45.6) Tier: 2 Group: 1 Technical
Reference:
1-AP-22.1 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
VIRGINIA POWER NORTH ANNA POWER STATION ABNORMAL PROCEDURE NUMBER PROCEDURE TITLE REVISION 1-AP-22.1 LOSS OF 1-FW-P-2 TURBINE-DRIVEN AFW PUMP 13 PAGE (WITH FOUR ATTACHMENTS) 1 of 6 PURPOSE To provide instructions for placing an AFW feed source in service when the Turbine-Driven AFW Pump is inoperable. ENTRY CONDITIONS This procedure is entered when AFW flow to the A SG is required and either of the following conditions exists:
- 1-FW-P-2 is inoperable, or
- Annunciator Panel F D-8, TURBINE DRIVEN AFW PUMP TROUBLE OR LUBE OIL TRBL, is LIT.
RECOMMENDED APPROVAL: DATE EFFECTIVE DATE RECOMMENDED APPROVAL - ON FILE APPROVAL: DATE APPROVAL - ON FILE
NUMBER PROCEDURE TITLE REVISION 13 1-AP-22.1 LOSS OF 1-FW-P-2 TURBINE-DRIVEN AFW PUMP PAGE 2 of 6 j STEP ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I CAUTION:
- When the ECST level decreases to 40%, then 1-AP-22.5, LOSS OF EMERGENCY CONDENSATE STORAGE TANK 1-CN-TK-1, should be initiated to provide an alternate water source to the AFW Pumps.
- To prevent heating of the ECST above 120°F, the AEW Pumps should not be run on recirc for extended periods of time.
- To prevent possible degradation of the AEW Pump. the amount of time spent on minimum recirc flow should be minimized.
NOTE:
- Normal PRZR spray should be isolated from any RCP that is stopped.
- The C RCP provides the best PRZR spray capability. The A RCP also provides PRZR spray capability.
- If AFW Pumps are lost due to loss of control from the Control Room, then evaluate using 1-AP-2O, OPERATION FROM THE AUXILIARY SHUTDOWN PANEL, to start the affected AFW Pumps from the Auxiliary Shutdown Panel.
l._ CHECK MAIN FEEDWATER - IN SERVICE Do the following: a) Stop all but one RCP. b) Initiate attempts to restore Main Feedwater.
- 2. CHECK BOTH MOTOR-DRIVEN AFW Start both Motor-Driven AEW Pumps:
PUMPS - RUNNING
- l-FW-P-3A
- 1-FW-P-3B
- 3. CHECK ECST LEVEL - GREATER THAN 40% Initiate 1-AP-22.5, LOSS OF EMERGENCY CONDENSATE STORAGE TANK 1-CN-TK-1.
NUMBER PROCEDURE TITLE REVISION 13 1-AP-22.1 LOSS OF ]-FW-P-2 TURBINE-DRIVEN AFW PUMP PAGE 3 of 6
STEP ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I NOTE:
- AFW HCVs fail open on loss of semi-vital bus 1A.
The AFW lineup drawings of Attachments 3 and 4 should be retained in the Control Room to provide Control Room personnel with a graphical representation of the AEW lineup. 4._ PERFORM ATTACHMENT 4 TO ALIGN THE Perform Attachment 3 to align the HCV HEADER FOR FEEDING ALL SGs MOV header for feeding all SGs. GO TO Step 7.
- 5. CONTROL AFW FLOW TO MAINTAIN SG NARROW RANGE LEVELS BETWEEN 23%
AND 50% USING: fJ
- 1-FW-HCV-100A for A SG
- 1-FW-HCV-100B for B SC
- 1-FW-HCV-100C for C SC U 6._ GO TO STEP 8 7._ CONTROL AFW FLOW TO MAINTAIN SG NARROW RANGE LEVELS BETWEEN 23%
AND 50% USING:
- 1-FW-MOV-100A for A SG
- 1-FW-MOV-100B for B SC
- 1FW-MOV-100C for C SG
- 8. DETERMINE IF AFW PUMPS CAN RETURN Continue with other procedures and TO NORMAL DISCHARGE ALIGNMENT steps in effect. WHEN normal alignment can be established, THEN GO TO Step 9.
9._ RAISE SC NARROW RANGE LEVELS TO 45% TO 50%
NUMBER ATTACHMENT TITLE REVISION l-AP-22.l 13 RETURNING AFW PUMPS TO NORMAL DISCHARGE ALIGNMENT ATTACHMENT PAGE 2 lof2 NOTE: Some of the AFW System valves have admin locks.
- 1. Locally close and lock the following valves (located in the Motor-Driven AEW Pumphouse):
* ]-EW-166, 3A Mtr Drvn AFW Pp to MOV Hdr Discharge Isol Valve
- 1-FW-62, AFW A MOV l-FW-MOV-100A Outlet Isolation Valve
- l-FW-l26, Aux Feedwater C MOV Outlet Isolation Valve
. 1-FW-l90, 3B Mtr Drvn AEW Pp to HCV Hdr Inlet Isol Valve
- 1-FW-64, AEW A HCV 1-FW-HCV-100A Outlet Isolation Valve
- 1-FW-96, AFW B HCV 1-FW-HCV-100B Outlet Isolation Valve
- 2. Notify the Control Room that Attachment 2 is complete and to return to 1-AP-22.1. step in effect.
I. H H C-) oSc z H Main Feed t::rj H z z C) hxj Service Water C) Drain U) H H 1 -FW-228 H C) 0 z Li Main Feed 0 H H C) H (I) C) 1 -FW- 164 1 -FW-PCV- 1 59A 1 -FW-P-3B C) en 1 -FW-MOV- 1 OOC 0 z z Main Feed z Drain -3 Fire Main AUXILIARY FEEDWATER SYSTEM NORMAL LINEUP tJ o F< 0 H,
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NUMBER ATTACHMENT TITLE REVISION 1-AP-22.1 13 ALIGNING BOTH MOTOR-DRIVEN AFW PUMPS TO HCV HEADER ATTACHMENT PAGE 4 lof2 NOTE: Some of the AFW System valves have admin locks.
- 1. Have the CR0 close the following AFW Valves:
. 1-FW-HCV-100A, AFW HCV HEADER TO A SC
. 1-FW-MOV-100D, TURBINE DRIVEN AEW PUMP TO A SC
- 1-FW-HCV-100B, AFW HCV HEADER TO B SC
. 1-FW-MOV-100B, AFW MOV HEADER TO B SC
- 2. Locally unlock and open the following valves (located in the Motor-Driven AFW Pumphouse)
. 1-FW-19O, 3B Mtr Drvn AFW Pp to HCV Hdr Inlet Isol Valve
. 1-FW-64. AFW A HCV 1-FW-HCV--100A Outlet Isolation Valve
. 1-FW-96, AEW B HCV 1-FW-HCV-100B Outlet Isolation Valve
- 3. Notify the Control Room that Attachment 4 is complete and to return to 1-AP-22.1, step in effect.
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QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 47. 062A1 .03 047/NEW//H/4/2.5/2.8/6/
Unit 1 is at 100% power. Which ONE of the following identifies the response if Vital Bus Inverter 1-Il fails, AND includes the Technical Specification implications of this malfunction? 12OVAC Vital bus 1-li will be powered automatically from A. MCC 1H1-1 via a constant voltage transformer; Unit 1 is in an info-only action. B. MCC 1H1-1 via a constant voltage transformer; Unit 1 is in a limiting action. C. MCC 1J1-1 via a constant voltage transformer; Unit 1 is in an info-only action. D. MCC 1J1-1 via a constant voltage transformer; Unit 1 is in a limiting action. A. Incorrect. First part is correct as this is the correct backup for this train of equipment. Second part is incorrect but plausible since the candidate may base his decision on TS 3.8.9 which would be sat since the bus is energized with proper voltage and frequency and overlook the additional requirements of TS 3.8.7. B. Correct. First part is correct as this is the correct backup for this train of equipment. Second part is also correct; as alluded to above TS 3.8.7 has additional requirements to those of 3.8.9 and the unit will be in a shutdown action statement. C. Incorrect. First part is incorrect but plausible since the candidate who lacks detailed system knowledge may be unsure of this and equate 1-Il with train 2 (or J train) equipment and erroneously select this distractor. Second part is also incorrect but plausible as discussed in distractor a. D. Incorrect. First part is incorrect but plausible as discussed in distractor c. Second part is correct as explained in answer b.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 AC Electrical Distribution System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including: (CFR: 41.5 / 45.5) Effect on instrumentation and controls of switching power supplies Tier: 2 Group: 1 Technical
Reference:
TS 3.8.7 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info: considered to meet intent of K/A; by design switching power supplies for a 12OVAC vital bus is a bumpless transfer, thus the true measurable effect on the instrumentation and controls is placing the unit in a limiting action due to the increased vulnerability to a loss of offsite power in this configuration where the affected components are reliant solely on the associated constant voltage transformer.
NUCLEAR DESIGN INFORMATION PORTAL InvertersOperati ng 3.8.7 3.8 ELECTRICAL POWER SYSTEMS 3.8.7 InvertersOperating LCO 3.8.7 The following inverters shall be OPERABLE.
- a. The Train H and Train J inverters; and
- b. The necessary inverters on the other unit for each required shared component.
- NOTE- -
One inverter may be disconnected from its associated DC bus for 24 hours to perform an equalizing charge on its associated battery, provided:
- a. The associated AC vital bus is energized from its constant voltage source transformer; and
- b. All other required AC vital buses are energized from their associated OPERABLE inverters.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One inverter required A.1 NOTE by LCO 3.8.7.a Enter applicable inoperable. Conditions and Required Actions of LCO 3.8.9, Di stri buti on SystemsOperati ng with any vital bus de-energi zed. Restore inverter to 7 days OPERABLE status. North Anna Units 1 and 2 3.8.7-1 Amendments 253/234
DOMINION 1 -OP-26A North Anna Power Station Revision 50 Page 19 of 141 (Page 3 of 3) Attachment 5 1-EE-MCC-IHI-1, IHI-1 Motor Control Center 1-EP-MC-1O LOCATION: EMERGENCY SWITCHGEAR ROOM POWER SUPPLY: I -EE-BKR 1 4H-4
REFERENCE:
1 1715-FE-1Z Breaker Required Ind No. Load Position Verifier Verifier 1-SW-P-i 0, In Service On G4 Rad Mon Sample Pump for I 1-SW-RM-108 Ckt Bkr (Ref. 2.3.26) Out of Service Off 1 -EP-IV-02, H1L On Computer Transfer Inverter Circuit Breaker (SOLA) 1 -EP-CB- 1 6A, H1R On 1A Semi-Vital Bus Distribution Pnl Ckt Bkr 1-EI-CB-41 B, H2L On Rod Position Indicator Rack No 2 Cabinet Ckt Bkr 1 -HV-FL-9, H2R On Filter Heater Coil Circuit Breaker 1-S W-MOV-i 19, H3 On Scm Wash Pps Mkup to SW Sply Hdr No 1 Isol Vv CB H4L SPARE, H4L Off Spare Circuit Breaker H4R On Vital Bus 1 -II Bypass Transformer Circuit Bkr
DOMINION 1-OP-26A North Anna Power Station Revision 50 Page 21 of 141 (Page 2 of 2) Attachment 6 1-EE-MCC-IHI-4, IHI-4 Motor Control Center 1-EP-MC-41 LOCATION: EMERGENCY SWITCHGEAR ROOM POWER SUPPLY: 1-EE-BKR-14H-2
REFERENCE:
I 1715-FE-1Z Breaker Required md No. Load Position Verifier Verifier C3L On Vital Bus 1-I Bypass Transformer Circuit Bkr C3R Spare, C3R Off Spare Circuit Breaker 1 -EP-DB- 1 6A, D1L On 1 6A Semi-Vital Bus Panel Circuit Breaker 1 -EP-DB-0 1, Normal Feed On MGPI Vent Stack RM Normal Feed Circuit Bkr Dl R Alternate Feed Off 1-BY-BC-02, Battery Charger 1-1 Power Supply Circuit D2L On Bkr I -BY-BC-04, Battery Charger 1 -II Power Supply Circuit D2R On Bkr 1-HV-E-4C, 4C Control Room Heat and Vent Chiller D3 On Circuit Bkr
- IF breaker is on with the Unit in Mode 5 or 6, THEN have the SRO enter into the Action Statement Status Log to open the Breaker prior to entering Mode 4. (Reference 2.1.2) c.J
DOMINION 1-OP-26A North Anna Power Station Revision 50 Page 27 of 141 (Page 3 of 4) Attachment 9 1-EE-MCC-IJI-1, IJI-1 Motor Control Center 1-EP-MC-11 LOCATION: EMERGENCY SWITCHGEAR ROOM _J POWER SUPPLY: 1-EE-BKR-14J-4
REFERENCE:
I 1715-FE-1P Breaker Required md No. Load Position Verifier Verifier Vital Bus 1 -III E1R On Bypass Transformer Circuit Breaker 1 -EP-PNL- 1 Cl, E2L On Control Room Lighting Panel Circuit Breaker 2-EP-CB- 129, E2R On Unit 2 Hydrogen Analyzer Alternate Feed Ckt Bkr 1 -SD-MO V-l OOC, E3 On HP Turbine C Main Stm Sply Hdr Drain Vv Ckt Bkr 1 -HV-F-57C, E4 On Service Building Battery Rm 1-Ill Sply Fan Ckt Bkr 1 -HV-AC-7, Fl On Emergency Switchgear Room A/C Unit 7 Circuit Bkr I -HV-AC- 1, F2 On Control Room A/C Unit I Circuit Breaker 1 -EP-DB-- 1 6B, F3L On 1 6B Semi-Vital Bus Panel Circuit Breaker NORMAL FEED Off 1 -EP-CB- 1 6C, MGPI Vent Stack Rrn Altn F3R ALTERNATE FEED On Power Supply Circuit Bkr F4 1-HV-F-57D, On Service Building Battery Room Exhaust Fan Ckt Bkr 1 -DB-P- 1 OB, F5 On 1 OB Chiller Room Sump Pump Circuit Breaker Vital Bus 1-TV Gl On Bypass Transformer Circuit Breaker
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 48. 062AA2.0 1 048/NEW//H/3/2.9/3 .5/4/
Given the following conditions:
- Both Units are at 100% power.
- Service Water (SW) is throttled.
- Service Water pumps 1-SW-P-lA and 1-SW-P-lB are running.
- All SW Spray MOVs are open and all SW Spray Bypass MOVs are closed.
The crew notes that SW reservoir level has been decreasing slowly, and currently indicates 312 feet. The following conditions exist:
- 1-SW-P-lA discharge pressure is 45psig.
- 1-SW-P-lB discharge pressure is 55 psig.
Based on these indications, the leak is on and the Ultimate Heat Sink is A. A Service Water Header; operable B. A Service Water Header; inoperable C. B Service Water Header; operable D. B Service Water Header; inoperable
- a. Incorrect. First part is correct; for the normal plant alignment of SW pumps coupled with the information that the system has been throttled the lower pressure associated with 1-SW-P-lA would be an indication of increased flow, which all things left unchanged would be indicative of a leak.
Second part incorrect but plausible as discussed in Distractor c.
- b. Correct. First part is correct as discussed in distractor a. Second part is correct the surveillance requirement for UHS operability (SR 3.7.9.1) is 313 feet.
- c. Incorrect. Plausible since the candidate may confuse which pump is on which header (they are different between the units), and/or may not recognize the significance of the information provided in determining leak location. Second part is plausible since step 1 of the procedure has the operator check level greater than 310 feet; the candidate who lacks detailed knowledge of Tech Specs may erroneously assume that this is the TS limit and therefore conclude that the UHS is still operable.
- d. Incorrect. First part is incorrect as discussed in Distractor c; second part is correct as discussed in distractor b.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Loss of Nuclear Service Water Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water: (CFR: 43.5/4513) Location of a leak in the SWS Tier: 1 Group: 1 Technical
Reference:
TS 3.7.9, O-AP-12, O-OP-49.6 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info: This is a Unit differences question. the SW pump header alignment is different between the two units.
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DolmoRl NORTH ANNA POWER STATION ABNORMAL PROCEDURE NUMBER PROCEDURE TITLE REVISION 36 0-A P-i 2 LOSS OF SERVICE WATER PAGE (WITH ELEVEN ATTACHMENTS) 1 of 17 PURPOSE To provide the instructions to use in the event of a loss of Service Water. ENTRY CONDITIONS This procedure is entered when there is a loss of Service Water as indicated by any or all of the following:
- Annunciator Panel J B-3, SERV WTR RETURN HDR LO FLOW, is LIT, or
- Annunciator Panel J D-3, SW PP 1-P1A, 2-P1A AUTO TRIP, is LIT, or
- Annunciator Panel J E-3, AUX SW PP 1-P4 LO FLOW, is LIT, or
- Annunciator Panel J H-3, SW PP 1 -P1 B, 2-Pi B AUTO TRIP, is LIT, or
- Annunciator Panel J D-4, AUX SW PP 2-P4 LO FLOW, is LIT, or
- Annunciator Panel J E-4, AUX SW PP 1-P4, 2-P4 AUTO TRIP, is LIT.
CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 36 O-AP-1 2 LOSS OF SERVICE WATER PAGE 2 of 17 H_STEP_[1 ACTION/EXPECTED RESPONSE j RESPONSE NOTOBTAINED NOTE:
- If a Reactor trip is initiated, then, to restore cooling to necessary plant equipment, this procedure should be performed in conjunction with Emergency Operating Procedures.
- Loss of Service Water will result in loss of Component Cooling and Residual Heat Removal Systems for the affected unit(s). This procedure should be performed in conjunction with the appropriate Abnormal Procedures.
- To ensure proper operation of important equipment, CC temperatures should be continuously monitored during loss of Service Water conditions.
- In emergency situations, Attachment 8, OPERATION OF AUXILIARY SERVICE WATER PUMPS may be used to place Auxiliary Service Water Pumps in service. If the Auxiliary Service Water pumps are placed in service, then NRC reportability requirements should be evaluated.
- If high volume blowdown of Service Water Reservoir is in service, then evaluate using O-OP-49.7, High Volume Blowdown Of The Service Water Reservoir, to secure blowdown.
- 1. CHECK SERVICE WATER SYSTEM INTEGRITY:
D a) Check Service Water Reservoir level - a) Do the following: GREATER THAN 310 FEET D 1) IF the Service Water Reservoir is intact, THEN begin a makeup using f) ( 1 either ATTACHMENT 10 or 0-OP-49.3, SERVICE WATER RESERVOIR MAKEUR D 2) IF level cannot be restored, THEN place the Auxiliary Service Water Pumps in operation using either ATTACHMENT 10 or 0-OP-49.2, SERVICE WATER SYSTEM LAKE TO-LAKE OPERATION. (STEP 1 CONTINUED ON NEXT PAGE)
DOMINION O-OP-49.6 North Anna Power Station Revision 19 Page 5 of 32 2.4.7 Plant Issue N-2006-0257, 1-CC-E-1A High SW zP Experienced When Flow Was Increased Due To Air In The Top Of The Heat Exchanger 2.4.8 CR002375, SW Blowdown Flow Adjustment 2.4.9 CA 139408; CR014429, Place keeper CR for PT N-2004-0272 mit Verif 3.0 INITIAL CONDITIONS Review the equipment status to verify station configuration will support the performance of this procedure. 4.0 PRECAUTIONS AND LIMITATIONS 4.1 Comply with the following guidelines when marking steps N/A:
- IF the conditional requirements of a step do not require the action to be performed, THEN mark the step N/A.
- IF any other step is marked N/A, THEN have the SRO approve the N/A and justify the N/A on the Procedure Routing Sheet.
4.2 Perform all operations in accordance with RWPs. Make every effort to maximize personnel safety while minimizing personnel exposure and area contamination, both surface and airborne. Use ALARA concepts. 4.3 The Service Water Bypass MOVs are in series and can be throttled. The normal method of throttling flow is to fully open the Unit 1 upstream valve and to throttle the Unit 2 downstream valve. Throttling of Service Water Bypass Valves CANNOT be used to satisfy the 54-psig throttling requirement. (Reference 2.4.3) 4.4 The Service Water System should be throttled in ALL Modes to at least 54 psig operating pump discharge pressure. (Reference 2.4.1) 4 alyic
DOMINION O-OP-49.6 North Anna Power Station Revision 19 Page 16of32 5.2 Throttling the Service Water System 5.2.1 Verify Initial Conditions are satisfied. 5.2.2 Review Precautions and Limitations. 5.2.3 Ensure the Service Water System is aligned to a throttled condition as follows:
- a. Ensure all of the following Service Water Spray Valves are open:
- 1-SW-MOV-121A, SPRAY 1A1 SW RESERVOIR
- 1-SW-MOV-121B, SPRAY 1B1 SW RESERVOIR
- 1-SW-MOV-122A, SPRAY 1A2 SW RESERVOIR
- 1-SW-MOV-122B, SPRAY 1B2 SW RESERVOIR
- 2-SW-MOV-221A, SPRAY 2A1 SW RESERVOIR
- 2-SW-MOV-221B, SPRAY 2B1 SW RESERVOIR
- 2-SW-MOV-222A, SPRAY 2A2 SW RESERVOIR
- 2-SW-MOV-222B, SPRAY 2B2 SW RESERVOIR
DOMINION O-OP-49.6 North Anna Power Station Revision 19 Page 17of32
- b. Ensure all of the following Service Water Spray Bypass Valves are closed:
- 1-SW-MOV-123A, HDR 4 BYPASS SW RESERVOIR
- 1.-SW-MOV-123B, HDR 3 BYPASS SW RESERVOIR
- 2-SW-MOV-223A, HDR 3 BYPASS SW RESERVOIR
- 2SW-MOV-223B, HDR 4 BYPASS SW RESERVOIR
- c. Ensure only one Service Water Pump is mnning on each Service Water Header.
DOMINION 0-OP-49.6 North Anna Power Station Revision 19 Page 18 of 32 CAUTION IF only one CCHX is in service on an individual header, THEN the z\P across the SW side of the CCHX should be less than 20 psid when operating on a SW header with one SW pump in service and another SW pump in auto. The CCHX AP may be increased to 25 psid when only one SW pump is operable on the header. (Reference 2.3.5)
- d. Throttle the Service Water Discharge Valves for the in-service CCHXs so that each in-service Service Water Pump discharge pressure, as is read in the Control Room, is at least 54 psig. Mark the remaining valves N/A.
(References 2.4.3 and 2.4.4)
- l-SW-230, (Unit 1) 1A Component Cooling HX SW Ret Hdr No 4 Isol Vv
- l-SW-241, (Unit 1) lB Component Cooling HX SW Ret Hdr No 4 Isol Vv
- 2-SW-184, (Unit 2) 1A Component Cooling HX SW Ret Hdr No 3 Isol Vv
- 2-SW-194, (Unit 2) lB Component Cooling HX SW Ret Hdr No 3 Isol Vv
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 49. 062K1 .04 049/BANK/NAPS/HJ3/3 .7/4.2/6/
Both Units are at 100% power. An electrical fault results in a loss of C RSST. Which ONE of the following will occur as a result of this event? A. 1H and 2J EDGs will start and load; breaker 15G10 will remain open B. 1H and 2J EDGs will start and load; breaker 15G10 will close C. ONLY 2H EDG will start and load; breaker 15G10 will remain open D. ONLY 2H EDG will start and load; breaker 15G10 will close
- a. Incorrect. First part is correct for the normal electrical lineup. Second part is incorrect but plausible since the candidate who lacks detailed systems knowledge may not adequately understand the design and features of this interlock function.
- b. Correct. First part correct as noted above. Second part is also correct, since the fault is associated with the RSST and not the associated G bus, the subject breaker is designed to auto-close.
- c. Incorrect. First part is incorrect but plausible as the candidate may lack conplete knowledge of the electrical system from RSSTs to transfer bus to emergency bus. Second part incorrect but plausible as discussed in Distractor a.
- d. Incorrect. First part is incorrect as discussed in Distractor c. Second part is correct as explained in answer b.
AC Electrical Distribution System Knowledge of the physical connections and/or cause effect relationships between the ac distribution system and the following systems: (CFR: 41.2 to 41.9) Off-site power sources Tier: 2 Group: 1 Technical
Reference:
0-AP-lO, att.2, AR H-G7 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
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VIRGINIA POWER 1-EI-CB-21H ANNUNCIATOR G7 1-AR-H-G7 NORTH ANNA POWER STATION REV. 0 APPROVAL: ON FILE Effective Date:05/09/97 4KV BUSSES 1G-2G TIE CLOSED 1.0 Probable Cause 1.1 T or C RSS transformer Loss of B 1.2 Fault on E or F transfer bus 1.3 1G and 2G Buses crosstied 2.0 Operator Action 2.1 Verify 15G10 closed. 2.2 Verify voltage on 1G and 2G buses. 2.3 Monitor amps on RSS Transformer carrying G buses. 2.4 IF 15G10 auto closed, THEN notify Electrical Department. 3.0 References 3.1 LSK221OC 3 2 LSK221OD 3.3 11715FE21K 4.0 Actuation 4.1 52A contact on 15G10
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 50. 063K4.04 050/BANK/NAPS/L/4/2.6/2.9/6/
Which ONE of the following describes the operation of the I25VDC Vital bus battery chargers? The battery chargers have an adjustable current limiter device, which A. opens the AC input breaker if AC input current reaches 275 amps. B. limits the maximum AC input current to 275 amps. C. opens the DC output breaker if DC output current reaches 275 amps. D. limits the maximum DC output current to 275 amps.
- a. Incorrect. Plausible since these are powered from emergency busses and the candidate who lacks detailed systems knowledge may consider this feature is for protection of the emergency bus.
- b. Incorrect. Plausible since these are powered from emergency busses and similar to the above the candidate who lacks detailed systems knowledge may consider this feature is for protection of the emergency bus.
- c. Incorrect. Plausible since the candidate who lacks detailed systems knowledge may consider this protection necessary to protect the charger from damage due to instantaneous over-current.
- d. Correct. This is correct, the feature is designed to allow the charger to meet the TS minimum capability of 270 amps, while restricting the maximum to within design limits.
D.C. Electrical Distribution Knowledge of DC electrical system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) Trips. Tier: Group: Technical
Reference:
1-PT-87.3H, SYSTEM DESIGN BASIS DOCUMENT Proposed references to be provided to applicants during examination: None Learning Objective: vision #5509 Question History: additional info:
PROCEDURE NO: Doniinion I -PT-87.3H REVISION NO: NORTH ANNA POWER STATION 15 PROCEDURE TYPE: UNIT NO: ELECTRICAL PERIODIC TEST 1 PROCEDURE TITLE: STATION BATTERY CHARGER I-I SERVICE TEST TEST FREQUENCY: UNIT CONDITIONS REQUIRING TEST: 18 months At all times SPECIAL CONDITIONS: None SURV REQ PMT REVISION
SUMMARY
FrameMaker Template Rev. 030. Feedback Incorporation Process Added 1-PT-87.1H to Step 6.2.45. REASON FOR TEST (CHECK APPROPRIATE BOX): Surveillance Post-Maintenance Work Order Number: TEST PERFORMED BY (SIGNATURE): DATE STARTED: DATE COMPLETED: TEST RESULT (CHECK APPROPRIATE BOX): CONDITION REPORT NUMBER(S) AND DATE: Satisfactory Unsatisfactory Partial THE FOLLOWING PROBLEM(S) WERE ENCOUNTERED AND CORRECTIVE ACTIONS TAKEN: (Use back for additional remarks.) COGNIZANT SUPERVISOR or DESIGNEE: DATE: ADDITIONAL REVIEWS: DATE: Battery Engineer: CONTINUOUS USE
DOMINION 1-PT-87.3H North Anna Power Station Revision 15 Page 13of31 NOTE: The battery charger current limiter is set for 275 amps. When the chargers output current demand exceeds the current limiter setting, the current limiter should respond by limiting output current to a value approximately equal to the current limiter setting. This happens when the charger is heavily loaded by connecting it to a discharged battery. If the current continues to increase above the current limiter setting when the charger is placed in service, the current limiter is NOT responding correctly. NOTE: Depending on the batterys state of charge when the charger is placed in service, the current limiter may or may NOT be challenged. If the charger current does NOT exceed 275 amps when the charger is restored to service, the current limiter is NOT tested properly. 6.2.18 Restore Station Battery Charger 1-I to service using l-OP-26.4.l and ensure OPS the NORMAL/EQUALIZE switch on Station Battery Charger 1-I is in the NORMAL position. 6 2 19 IF the chargers output current was between 275 and 285 amps when the charger was restored to service, THEN mark N/A Steps 6.2.20 through 6.2.30. 6.2.20 Energize the Battery Analyzer. 6.2.2 1 Verify the test equipment is functioning normally. 6.2.22 Program the Battery Analyzer as follows:
- Test Duration 1 hour
- Load Bank Load 50 amps
- Low Battery Voltage Alarm 125 volts
- Low Battery Voltage Shutdown 0 volts
- Low Cell Alarm 0 volts
- Low Cell Shutdown 0 volts
125V DC EMERGENCY POWER SYSTEM NORTH ANNA POWER STATION Domhiiew SYSTEM DESIGN BASIS DOCUMENT Table 4.2-1 Safety-Related Performance Criteria Req u rement Design Feature
- 1. In support of the function in Table 4.1-ito The ED System batteries are specified to have a supply and distribute Class 1 E DC power, the current capacity of 1210 AH that is greater than ED System shall provide sufficient capacity to the required values given in Table 5.1-i. [11.1.11]
power the required Class 1 E DC loads for the [11.3.3] battery duty cycle without an external power source and without automatically shedding non-Class iE loads. [11.4.22] [11.7.1] [11.10.11
- 2. In support of the function in Table 4.1-i to The ED System battery chargers are specified to supply and distribute Class 1 E DC power, the be capable of providing 250A continuous with a ED System shall provide sufficient capacity to current limiting action starting at 275A that is power the required Class 1 E DC loads while greater than the required values given in recharging the ED System batteries within a Table 5.1-1. [11.1.14] [11.3.2] [11.9.11] [11.9.24]
minimum time period without automatically [11.7.20] shedding non-Class 1 E loads. [11.4.22] [li.7.i][li.iO.1}
- 3. In support of the function in Table 4.1-ito The ED System 125 VDC emergency supply and distribute Class 1 E DC power, the switchboards are specified to be capable of ED System equipment shall provide sufficient withstanding a short circuit current of 20,000A capability to withstand the highest available that is greater than the required values given in short circuit current without automatically Table 5.1-i. [11.1.12] [11.1.13] [11.3.1]
shedding non-Class 1 E loads. [11.4.22] [11.7.1] [11.10.1]
- 4. In support of the function in Table 4.1-ito The ED System supplied equipment is specified supply and distribute Class 1 E DC power, the to be capable of operating at a voltage of i4OV voltage of the ED System shall be maintained that is the same value as the required value given within the required voltage range during in Table 5.1-1. The ED System battery chargers operation without automatically shedding are supplied with controls to adjust the maximum non-Class iE loads. [11.4.22] [11.7.1] float and equalizing voltage settings. A trouble
[11.10.1] alarm for the battery chargers is provided in the Control Room. [11.1.12] [11.1.10] [11.3.2] [11.2.11] [11.2.37] [11.4.43] The ED System supplied equipment is specified to be capable of operating at a voltage of iO5V that is the same value as the required values given in Table 5.1-i. The ED System batteries are capable of providing the required minimum voltage to the 125 VDC emergency buses. [11.1.11] [11.3.3] [11.4.43] SDBD-NAPS-ED PROPRIETARY REVISION NO. 11 CHO4N.FM EFFECTIVE DATE: 08/24/09 4-6
STUDENT GUIDE FOR VITAL AND EMERGENCY ELECTRICAL DISTRIBUTION SYSTEM (35) Vital Electrical Distribution Topic 11 125-volt DC battery chargers 1.1 Objective U 5509 List the following information associated with the 125-volt DC battery chargers.
- Purpose
- Source of power to each normal and swing battery charger (SOER-81-15)
- Color designation associated with each vital DC bus
- Output voltage during normal operation (SOER-81-15)
- Output voltage during an equalizer battery charge
- Nominal output current (SOER-81 -1 5)
- Maximum current flow (SOER-81 -15) 1.1 Content
- 1. The battery chargers convert 480 VAC power to a 125 VDC regulated output, which powers the associated 125 VDC buses and maintains a floating charge on the batteries connected to the buses.
- 2. The normal and swing battery chargers are powered from their respective 480 volt emergency busses:
1.1. 1H1-4 supplies input power to normal battery chargers 1-I and 1-Il and swing battery charger iC-I 1.2. 1J1-i supplies input power to normal battery chargers 1-Ill and 1-lV and swing battery charger iC-Il
- 3. Each vital DC bus has a designated identifying color:
1 .3. Red identifies the I-i vital bus. 1 .4. White identifies the 1-2 vital bus 1 .5. Blue identifies the 1-3 vital bus 1 .6. Yellow identifies the 1-4 vital bus REACTOR OPERATOR Page 4 of 70 Revision 6
STUDENT GUIDE FOR VITAL AND EMERGENCY ELECTRICAL DISTRIBUTION SYSTEM (35)
- 4. During normal operation, the battery output voltage is 135 volts
- 5. During an equalizer battery charge, the battery output voltage is 1 39.8 volts
- 6. During normal operation, battery charger 1-I (2-1) have an approximate current output of 105 amps.
1.7. Battery chargers 1-Il, 1-Ill, and 1-IV have an approximately current output of 60 amps. 1 .8. The battery chargers are capable of providing 250 amps continuously.
- 7. The maximum current flow is limited to 275 amps by an adjustable current limiter.
Topic 1 2 Kirk Interlock 1.2 Objective U 5510 Explain how the Kirk interlock ensures that only one DC bus is supplied from the swing battery charger at one time (SOER-81 -1 5). 1.2 Content
- 1. A single key is used for two locking mechanisms, that cant be removed unless the breaker is open.
1.1. This ensures that only one DC bus is supplied from the swing battery charger at one time. Topic 1.3 120-volt DC Bus Breaker Panels 1.3 Objective U 4745 List the following information associated with the 120-volt DC bus breaker panels (SOER-81 -15).
- Location of each DC bus breaker panel
- Indication used to detect a ground on the panel
- Means provided to identify which DC load has a ground REACTOR OPERATOR Page 5 of 70 Revision 6
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 51. 064A4. 10 05 1/BANKJNAPS ID:2409/L/2/3 .3/3.4/6/
With Unit 1 at 100% power, the 1H 4160V bus normal feeder breaker 15H11 trips open, and 1H EDG re-energizes the bus. Per 0-AP-1 0, Loss of Electrical Power, configuring 1 H bus loads is done as a high priority in order to A. prevent EDG overload in the event of an SI/CDA. B. facilitate restoration of the emergency bus from the normal feeder. C. ensure loads that may have automatically started are returned to their pre-event (standby) status. D. ensure that 1 H bus loads will trip selectively in the event of an Sl/CDA.
- a. Correct. Per 0-AP-1 0, this action is necessary to ensure that when the EDG is the sole source, it will not be overloaded in the event of a DBA.
- b. Incorrect. Plausible since the candidate who has not internalized the true reason may conclude that it is a matter of course, since actions will be underway to restore the normal power source if possible.
- c. Incorrect. Plausible since auto starting of some equipment will occur. The associated attachment for configuring loads however is not designed for this purpose, and in fact may prefer to have a load that auto-started be left running, but again if the candidate does not fully understand how the attachment works may default to this distractor as it appears logical on the surface.
- d. Incorrect. Plausible because the candidate who does not fully understand load shed and/or sequencing might conclude that this is the correct answer since proper operation of those functions is important, however as discussed above it is not the reason (purpose) for performing the attachment and the attachment in itself will not control selective tripping of components.
Emergency Diesel Generator (ED/G) System Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8) Need for, and consequences of, manually shedding (loads) safeguards bus Tier: 2 Group: 1 Technical
Reference:
0-AP-lO Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
NUMBER PROCEDURE TITLE REVISION 62 O-AP-1 o LOSS OF ELECTRICAL POWER PAGE 15 of 32 -J_STEP J ACTION! EXPECTED RESPONSE J RESPONSE NOT OBTAINED
- 32. CHECK UNIT 2 120-VAC SEMI-VITAL BUS -
ENERGIZED: D a) 2A - VOLTAGE INDICATED C a) Initiate ATTACHMENT 7. C b) 2B - VOLTAGE INDICATED C b) Initiate ATTACHMENT 6. NOTE: If an EDG is the sole source of power to its Emergency Bus, then initiating ATTACHMENT 21, UNIT 1 EDG LOAD CONFIGURATION TO PREVENT OVERLOADING should be a high priority to prevent overloading of the EDO.
- 33. VERIFYALIGNMENTOFUNIT1 EMERGENCY DIESEL GENERATORS:
\
C a) Verify ANY Unit 1 Emergency Diesel Generator THE SOLE SOURCE OF I
) C a) GO TO Step 34.
POWER TO THE RESPECTIVE / ENERGIZED EMERGENCY BUS
/
C b) SI OR CDA has been ACTUATED
/
/ C b) Initiate ATTACHMENT 21, UNIT 1 EDO LOAD CONFIGURATION TO PREVENT OVERLOADING r
NUMBER ATTACHMENT TITLE ATTACHMENT 0-AP-lO 21 UNIT 1 EDG LOAD CONFIGURATION TO PREVENT OVERLOADING REVISION PAGE 62 iof6 NOTE: Monitor EDG load parameters closely during any pump operations.
- 1. IF 1 H EDG is the sole source of power to 1 H Emergency Bus, THEN do the following to limit the amount of instantaneous loading that could occur in the event of an SI/CDA:
a) Start 1-SW-P-iA. b) IF 2-SW-P-i B is running on the same SW Header as 1-SW-P-iA AND 2J EDG is NOT the sole source of power to the 2J Emergency bus, THEN 2-SW-P-i B may be placed in AUTO-AFTER-STOP NOTE: 1-CH-P-1C, C Charging Pump, has no automatic start features. If both 1-CH-P-1A and 1-CH-P-iB are running, then 1-CH-P-1C will receive an auto trip. c) Align Charging Pumps as follows:
- 1) IF 1-CH-P-iC is running on the 1H bus, THEN place i-CH-P-iA in Pull-To-Lock.
- 2) IF i-CH-P-1C AND i-CH-P-1B are BOTH running on the 1J bus, THEN place i-CH-P-1C in AFTER-STOP
- 3) IF 1 -CH-P-1 C is NOT running on the 1 H bus AND 1 -CH-P-1 A is available, THEN start 1 -CH-P-1 A.
- 4) IF i-CH-P-1B is running, AND 1J EDG is NOT the sole source of power to the iJ Emergency Bus, THEN 1-CH-P-1B may be placed in AUTO-AFTER-STOP
- 5) HZ 1 -CH-P-i B was placed in AUTO-AFTER-STOP, THEN clear the Lockout Alarm on 1 -CH-P-1 C.
(STEP 1 CONTINUED ON NEXT PAGE) V
NUMBER ATTACHMENT TITLE ATTACHMENT 0-AP-lO 21 UNIT 1 EDG LOAD CONFIGURATION TO PREVENT OVERLOADING REVISION PAGE 62 2of6 IF 1 H EDG is the sole source of power to 1 H Emergency Bus, THEN do the following to limit the amount of instantaneous loading that could occur in the event of an SI/CDA: (Continued) NOTE: The basis for the EDG load limit is 1500 Kw of loads will start on a CDA. The EDG is rated for 3000 Kw. The EDG cannot initially be loaded >1500 Kw, with the exception of loads that will trip on a CDA signal and the load of a running AFW pump. d) 1H EDG loading prior to any accident must be limited to 1500 Kw plus the running load values listed below. Determine 1 H EDG load limit as follows: (x) Total of KW ratings of all RUNNING equipment from the table below.
+l500Kw=
(x) EDG Load Limit RUNNING LOADS KW 1-CC-P-lA 311 1 -HV-F-37A 29 1-HV-F-37B 29 1 -HV-F-37C 29 1-HV-F-1A 155 1-HV-F-1C (If Running on 1H 155 Bus) 1-FW-P-3A 284 e) IF existing EDG load is greater than the limit calculated in Step 1 d above, THEN based on plant conditions, the equipment listed below may be secured to reduce EDG load to less than or equal to the calculated limit. Refer to KW Ratings For Individual Loads table on the last page of this attachment, as necessary: LOADS
. 1-HV-E-4A
- 1-IA-C-i
- i-HV-F-8A
- 1-HV-E-4C
- 1-CV-P-3A
- Group 4 PZR Heaters
- 1-CW-P-2B
- 1-FC-P-1A
- 1-HV-AC-2
- 1-HV-AC-6 (STEP 1 CONTINUED ON NEXT PAGE)
NUMBER ATTACHMENT TITLE ATTACHMENT 0-AP-lO 21 UNIT 1 EDO LOAD CONFIGURATION TO PREVENT OVERLOADING REVISION PAGE 62 3of6 IF 1 H EDG is the sole source of power to 1 H Emergency Bus, THEN do the following to limit the amount of instantaneous loading that could occur in the event of an SI/CDA: (Continued) f) if EDG load was reduced to less than or equal to the calculated limit and is not paralleled to another bus, THEN TS 3.8.1 Condition B, H, and K may be exited and Condition A still applies.
- 2. IF 1J EDO is the sole source of power to 1J Emergency Bus, THEN do the following to limit the amount of instantaneous loading that could occur in the event of an Sl/CDA:
a) Start 1-SW-P-lB. b) IF 2-SW-P-i A is running on the same SW Header as 1-SW-P-i B AND 2H EDG is NOT the sole source of power to the 2H Emergency bus, THEN 2-SW-P-lA may be placed in AUTO-AFTER-STOP c) Align Charging Pumps as follows:
- 1) IF 1-CH-P-1C is running on the 1J bus, THEN place 1-CH-P-1B in Pull-to-Lock.
- 2) H 1 -CH-P-1 C AND 1 -CH-P-1 A are BOTH running on the 1 H bus, THEN place 1 -CH-P-1 C in AFTER-STOP
- 3) IF 1-CH-P-1C is NOT running on the 1J bus AND 1-CH-P-1B is available, THEN start 1 -CH-P-i B.
- 4) IF i-CH-P-1A is running, AND 1 H EDG is NOT the sole source of power to the 1 H Emergency Bus, THEN i-CH-P-iA may be placed in AUTO-AFTER-STOP
- 5) H i-CH-P-1A was placed in AUTO-AFTER-STOP THEN clear the Lockout Alarm on 1 -CH-P-i C.
(STEP 2 CONTINUED ON NEXT PAGE)
NUMBER ATTACHMENT TITLE ATTACHMENT 0-AP-lO 21 UNIT 1 EDG LOAD CONFIGURATION TO PREVENT OVERLOADING REVISION PAGE 62 4of6
- 2. IF 1J EDG is the sole source of power to 1J Emergency Bus, THEN do the following to limit the amount of instantaneous loading that could occur in the event of an SI/CDA: (Continued)
NOTE: The basis for the EDG load limit is 1500 Kw of loads will start on a CDA. The EDG is rated for 3000 Kw. The EDG cannot initially be loaded >1500 Kw, with the exception of loads that will trip on a CDA signal and the load of a running AFW pump. d) 1J EDG loading prior to any accident must be limited to 1500 Kw plus the running load values listed below. Determine 1J EDG load limit as follows: (x) Total of Kw ratings of all RUNNING equipment from the table below.
+l500Kw=
(x) EDG Load Limit RUNNING LOADS KW 1-CC-P-lB 311 1-HV-F-37D 29 1-HV-F-37E 29 1-HV-F-37F 29 1-HV-F-1B 155 1 -HV-F-1 C (If Running on 1J 155 Bus) 1 -FW-P-3B 284 e) IF existing EDG load is greater than the limit calculated in Step 2d above, THEN based on plant conditions, the equipment listed below may be secured to reduce EDG load to less than or equal to the calculated limit. Refer to KW Ratings For Individual Loads table on the last page of this attachment, as necessary: LOADS
- 1-HV-E-4B
- 1-CV-P-3B
- 1-HV-F-8B
- 1-HV-AC-1
- 1 -HV-AC-7
- Group 1 PZR Heaters (STEP 2 CONTINUED ON NEXT PAGE)
NUMBER ATTACHMENT TITLE ATTACHMENT 0-AP-lO 21 UNIT 1 EDG LOAD CONFIGURATION TO PREVENT OVERLOADING REVISION PAGE 62 5of6
- 2. IF 1J EDG is the sole source of power to 1J Emergency Bus, THEN do the following to limit the amount of instantaneous loading that could occur in the event of an SI/CDA: (Continued) f) if EDG load was reduced to less than or equal to the calculated limit and is not paralleled to another bus, THEN TS 3.8.1 Condition B, H, and K may be exited and Condition A still applies.
- 3. WHEN Offsite Power is restored to the Unit 1 Emergency Busses, THEN restore the Service Water and Charging systems to normal as follows:
a) Restore Service Water Pump alignment as directed by the SRO using O-OP-49.4, SHIFTING SERVICE WATER COMPONENTS (PUMPS AND SPRAYS). Ensure all operable SW Pumps not running are in AUTO. b) Restore Charging Pump alignment as directed by the SRO using 1 -OP-8.1, CHEMICAL AND VOLUME CONTROL SYSTEM. Ensure all operable Charging Pumps not running are in AUTO-AFTER-STOR
NUMBER ATTACHMENT TITLE ATTACHMENT 0-AP-lO 21 UNIT 1 EDG LOAD CONFIGURATION TO PREVENT OVERLOADING REVISION PAGE 62 6of6 KW Ratings For Individual Loads 4160 VOLT LOADS MARK NUMBERS KW Charging Pump 1(2)-CH-P-1A, B, C 775 Component Cooling Pump 1(2)-CC-P-iA, B 450 Service Water Pump 1(2)-SW-P-iA, 1B,4 375 Auxiliary Feedwater Pump 1 (2)-FW-P-3A, B 336 Outside Recirc Spray Pump 1(2)-RS-P-2A, B 300 RHR Pump 1(2)-RH-P-iA, B 225 Low Head SI Pump 1(2)-SI-P-lA, B 188 480 VOLT LOADS MARK NUMBERS KW Inside Recirc Spray Pump 1(2)-RS-P-1A, lB 225 Quench Spray Pump 1(2)-QS-P-1A, B 188 Containment Recirc Fan 1(2)-HV-F-1A, B, C 150 PRZR Heater, Backup Group 1 268 nominal PRZR Heater, Backup Group 4 268 nominal MAJOR 480 VOLT MCC LOADS MARK NUMBERS KW Instrument Air Compressor and Pump 1(2)-IA-C-i and 1(2)-IA-P-i 100 Spent Fuel Pit Cooling Pump 1-FC-P-1B, iA 75 Auxiliary Building Central Exhaust Fan 1-HV-F- 8A, 8B, 8C 30 CRDM Cooling Fan 1 (2)-HV-F-37A, B, C, D, E, F 57 Bearing Lift Pump 1(2)-GM-P-b 22 Turning Gear Oil Pump 1 (2)-TM-P-i 45 SFGDS Exhaust Fan 1(2)-HV-F-40A, B 19 Turning Gear 1(2)-TM-P-2 37 Boric Acid Transfer Pump 1-CH-P-2 A, B, C, D 12 Containment Air Compressor i(2)-IA-C-2A, B 15 Chiller and Associated Pumps 1 (2)-HV-E-4A, B, C 93 Primary Grade Water Standby Pump 1-PG-P-2A, B 30 Auxiliary Building App R Fan 1-HV-F-75A, B 30
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 52. 065AG2.1 .27 O52INEW//L13/3.9/4.0181 Which ONE of the following identifies ALL of the Auxiliary Feedwater System components that are equipped with seismic air tanks?
A. AFW Pump 1-FW-P-2 steam supply valves (1-MS-TV-lilA & B) ONLY. B. AFW HCVs (1-FW-HCV-100A, B, & C) AND AFW PCVs (l-FW-PCV-159A & B) ONLY. C. AFW Pump 1-FW-P-2 steam supply valves (1-MS-TV-il 1A & B) AND AFW HCVs (l-FW-HCV-100A, B, & C) ONLY. D. AFW HCVs (l-FW-HCV-IOOA, B, & C) AND AFW PCVs (i-FW-PCV-159A & B) AND AFW Pump 1-FW-P-2 steam supply valves (i-MS-TV-111A& B).
- a. Incorrect. Plausible since the AFW HCVs fail open on loss of instrument air and thus would fulfill the Safety Related function of supplying the minimum required AFW flow in the required time under all postulated accident conditions; thus the candidate who lacks detailed systems knowledge may apply this mindset and conclude that the other choices are merely distracters and select this response.
- b. Incorrect. Plausible since again these valves are directly in the feed flow path; thus the candidate who lacks detailed systems knowledge may conclude that they would be the only ones that need a backup because of the failure mode (open) of the l-FW-P-2 pump steam supply valves.
- c. Incorrect. Plausible because the candidate who lacks detailed system knowledge may conclude that for the postulated scenario this minimum equipment is all that is necessary to stabilize and control the plant. While this is true, the actual plant design is more robust.
- d. Correct. All of the listed components are provided with this feature as described in the TS bases.
Loss of Instrument Air Knowledge of system purpose and/or function. (CFR: 41.7) Tier: 1 Group: 1 Technical
Reference:
AFW system design basis document, dwgs 1171 5-FM-082M sh. 1 & 082N Sh. 2 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
AUXILIARY FEEDWATER SYSTEM NORTH ANNA POWER STATION Dominion SYSTEM DESIGN BASIS DOCUMENT control valves i-FW-PCV-159A and B located on the pump discharge. [11.2.37] [11.2.107] [11.2.38] [11.2.108] During normal plant operation, the AFW System is in the standby mode with each pump dedicated to a particular steam generator resulting in three independent flow paths. The dedicated line-up ensures the required flow from one AFW pump to a single steam generator initially, and, assuming operator action within 30 minutes, from two AFW pumps to two steam generators after 30 minutes. [11.1.24] [11.1.54] The analysis assumed the turbine-driven pump flow was lost out of the break, one of the two motor-driven pumps was lost due to a single failure, and the remaining motor-driven pump fed its dedicated steam generator. Also, cross-connect valves on the pump discharge lines allows any AFW pump to be aligned to any steam generator, to allow feeding two steam generators within 30 minutes. [11.2.37] [11.2.107] [11 .2.38] [11.2.108] The flow control valves can be remotely controlled from the main control board or from the auxiliary shutdown panel. [11.2.100] [11.2.99] [11.2.9] [11.2.10] [11.2.11] Air bottles located in the main steam valve house (MSVH) provide a backup supply of instrument Instrument and Service Air air for the steam supply valves 1-MS-TV-ill A and B during loss of the (IA) System. Air bottles located in the motor-driven auxiliary feedwater (MDAFW) pumphouse provide backup air supply for i-FW-HCV-iOOA, B, and C and i-FW-PCV-i59A and B. These bottles are sized to provide an operator with the capability to manually operate essential valves and ensure thatAFW System flow requirements are maintained. [11.2.39][11.2.109][11.2.40] 2.2 SYSTEM OPERATING CONDITIONS This section describes the operation of the AFW System under various system operating conditions. The operating conditions are grouped by the ANSI N18.2 categories, i.e., Conditions I through IV events, as well as non-safety-related with special quality/regulatory requirements (NSQ) events that are not covered by the ANSI classification system. The categorization of operating conditions in this manner ensures consistency with other chapters of the SDBD. 2.2.1 Condition I Events Normal Operation and Operational Transients Safe Shutdown Maintenance The AFW pumps are started manually to perform safe shutdown maintenance. Steam generator levels are controlled manually by the Control Room operators. During other Condition I events, the following conditions shall be maintained when the reactor coolant is above 350°F and 450 psig [11.7.39]:
- 1. The temperature of the water contained in the ECST shall be maintained within limits established by the plant safety analysis or other physical design limitations imposed on the system. The volume of water contained in the ECST shall be maintained greater than limits established by system calculation. [11 .1.70]
SDBD-NAPS-AFW PROPRIETARY REVISION NO. 10 CHO2N.FM EFFECTIVE DATE: 08/11/09 2-2
ITO DAMPER FOR SAFEGUARD MACA FANS) 51 I90- Md-Slut 038 GN ONIRRUD I 6 4 2 lLEAl5 NOOES I. FOR ADDITIONAL NOTES SEE SREEO 1 OF IITIS-FN-082A.
- 2. FIELD ICONNORVCOIONI TB LECATE.FABRICATE & INSTALL SEISMIC RANGERS.
- 3. ORE ROVOING OF ORESE PIPES TO CONFORM TO ORE HIGH ENERGY PIPE/WHIP CRITERIA.
FIELD TB INSTALL.
- 4. NUMBERS ALLOCAOEO 00 0015 SREEO 521-599
[i 3/U-ACC-G5-2IB-S I LEVEL 15 3/4-ACCE1-21B-S I 5. LASS NLRIBER USED 978 LEVEL 5 I3,0-A-522S I LEVEL IS S. NUMBERS AVAILABLE 979-990 LJ 1 VCF 7. GRADE LEVELS FOR II7I5-FM-082M SR. I ONLY lODE LEVEL 15 = EL.272 0 MAIM 50EV)) AAVLE MUVSE I 3/R-VCC-95-2IB-S
) 3 ANY. FEEORATEB PUMP HOUSE OVENCA SPRAY PUMP MOUSE LEVEL IS EL.256-0 OUESCR SPRAY POMP ROUSE.
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QUESTIONS REPORT for NAPS 2010 RD NRC Exam rev3
- 53. 068K6. 10 053/NEW//L/3/2.5/2.9/9/
Which ONE of the following describes the impact of a failure of i-RM-LW-1i0, Clarifier Inlet Radiation Monitor, and 1 -RM-LW-1 11, Clarifier Outlet Radiation Monitor, on the following valves? 1-LW-PCV-1 15, LW Effluent to Discharge Canal 1-LW-FCV-100, Holdup Tank Influent A failure (HIGH) of A. either i-RM-LW-1iO or i-RM-LW-1 11 will close 1-LW-PCV-115 ONLY. B. either 1 -RM-LW-1 10 or 1 -RM-LW-1 11 will close 1 -LW-PCV-i 15 and 1 -LW-FCV-1 00. C. jjy i-RM-LW-111 will close 1-LW-PCV-115 ONLY. D. jjy 1 -RM-LW-i ii will close 1-LW-PCV-i 15 1 -LW-FCV-1 00.
- a. Incorrect. Plausible since the candidate who lacks detailed systems knowledge would see this as logical and thus likely default to it; also closing i-LW-PCV-i 15 will stop releases to the environment and procedures that stop releases only address this valve.
- b. Incorrect. Plausible as noted above; the second part is correct the FCV also has an auto close feature.
- c. Incorrect. First part is correct only the outlet monitor has auto closure feature for the subject valves, other than alarm function, LW-i 10 has no automatic functions. Also as noted above both valves get closed, not just the PCV.
- d. Correct. First part is correct as discussed in Distractor c. Second part is also correct as discussed above.
Liquid Radwaste System (LRS) Knowledge of the effect of a loss or malfunction on the following will have on the Liquid Radwaste System: (CFR: 41.7 /457) Radiation monitors Tier: 2 Group: 2 Technical
Reference:
0-AP-5.i Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
NUMBER PROCEDURE TITLE REVISION 28 O-AP-5 1 COMMON UNIT RADIATION MONITORING SYSTEM PAGE 3 of 4 H_STEP ACTION/EXPECTEDRESPONSE I I RESPONSENOTOBTAINED 3.. VERIFY THE FOLLOWING FOR THE D Initiate the appropriate attachment(s) listed in the AFFECTED RADIATION following table(s). MONITOR(S): D
- Indication - NORMAL D
- Trend Recorder NORMAL E1
- Switch positions NORMAL
- Alarms NOT LIT INIT RADIATION MONITOR RECORDER AU. NO.
DECONTAMINATION BUILDING 1 -RM-RR-1 50 ATTACHMENT 8 1 -RMS-RM-1 51 AUX BUILDING CONTROL AREA 1-RM-RR-150 ATTACHMENT 9 1 -RMS-RM-1 54 WASTE SOLIDS AREA 1-RM-RR-150 ATTACHMENT 8 1 -RMS-RM-1 55 SAMPLE ROOM 1-RM-RR-150 ATTACHMENT 10 1 -RMS-RM-1 56 CONTROL ROOM 1-RM-RR-150 ATTACHMENT 11 1 -RMS-RM-1 57 LABORATORY AREA 1-RM-RR-150 ATTACHMENT 10 1 -RMS-RM-1 58 CLARIFIER INLET 1-RM-RR-150 ATTACHMENT 4 1-RM-LW-1 10 CLARIFIER OUTLET 1-RM-RR-15O ATTACHMENT 5 1-RM-LW-1 11 COMPONENT COOLING WATER 1-RM-RR-150 ATTACHMENT 7 1 -RM-CC-1 20 (STEP 3 CONTINUED ON NEXT PAGE)
NUMBER ATTACHMENT TITLE ATTACHMENT O-AP-5.1 4 CLARIFIER INLET RADIATION MONITOR REVISION PAGE 28 lof2
- 1. Notify Health Physics to obtain and analyze a sample from the Clarifier lnfluent.
- 2. IF the abnormality of the Radiation Monitor was caused by an obvious malfunction, THEN do the following:
a) IF Radiation Monitor has failed high AND reset is desired by removing fuses, THEN do the following:
- 1) Remove Radiation Monitor fuses.
- 2) Install Radiation Monitor fuses.
NOTE: If the Radiation Monitor resets, then a Work Order is not required. The Condition Report will be assigned to the Radiation Monitor Engineer for trending and evaluation.
- 3) IF the Radiation Monitor resets, THEN do the following. IF NOT, THEN GO TO Step 2b:
- Enter a Condition Report.
RETURN TO O-AP-5.1, COMMON UNIT RADIATION MONITORING SYSTEM, Step in effect. b) Inform the HP Shift Supervisor of the date and time the monitor was declared inoperable. c) Enter a Condition Report. d) RETURN TO O-AP-5.1, COMMON UNIT RADIATION MONITORING SYSTEM, Step in effect. NOTE: The Clarifier Effluent Radiation Monitor, which provides indication for a release pathway to the environment, has a higher priority than the Clarifier Influent Radiation Monitor.
- 3. Stop any Liquid Waste release in progress.
- 4. Observe activity level on 1 -LW-RM-1 11, Clarifier Effluent Radiation Monitor.
- 5. Determine the source of radiation as follows:
a) Check proper fill lineup for Liquid Waste tanks. b) Check proper recirc lineup for Liquid Waste tanks. c) Check proper discharge lineup for Liquid Waste tanks. d) Resample and analyze Liquid Waste tanks were that being discharged.
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NUMBER ATTACHMENT TITLE ATTACHMENT O-AP-5.1 5 CLARIFIER OUTLET RADIATION MONITOR REVISION PAGE 28 lof3 NOTE: The analysis should be completed immediately due to time limits for event classification in EPIP-1 .01, EMERGENCY MANAGER CONTROLLING PROCEDURE. NOTE: VPAP-2103N Action requirements may apply.
- 1. Notify Health Physics to sample the Clarifier Effluent.
- 2. IF the abnormality of the Radiation Monitor was caused by an obvious malfunction, THEN do the following:
a) IF Radiation Monitor has failed high AND reset is desired by removing fuses, THEN do the following:
- 1) Remove Radiation Monitor fuses.
- 2) Install Radiation Monitor fuses.
NOTE: If the Radiation Monitor resets, then a Work Order is not required. The Condition Report will be assigned to the Radiation Monitor Engineer for trending and evaluation.
- 3) IF the Radiation Monitor resets, THEN do the following. IF NOT, THEN GO TO Step 2b:
- a. Check equipment status:
- Ensure 1-LW-PCV-115 is open.
- Ensure 1-LW-FCV-100, HOLDUP TANK INFLUENT is open.
IF SG Blowdown pumps were running, THEN restart pumps, as required.
- b. Do the following:
- Enter a Condition Report.
- RETURN TO 0-AP-5.1, COMMON UNIT RADIATION MONITORING SYSTEM, Step in effect.
b) Inform the HP Shift Supervisor of the date and time the monitor was declared inoperable. c) Enter a Condition Report. d) RETURN TO 0-AP-5.1, COMMON UNIT RADIATION MONITORING SYSTEM, Step in effect.
NUMBER ATTACHMENT TITLE ATTACHMENT O-AP-5.1 5 CLARIFIER OUTLET RADIATION MONITOR REVISION PAGE 28 2of3 NOTE: Unit 1 and Unit 2 SG Blowdown Radiation Monitors are normally lined up to 1-LW-TK-13, Clarifier Hold-up Tank. If a Units SG Blowdown Radiation Monitors are lined up to the respective SG Blowdown Tank, then those Radiation Monitors will not be available following a complete isolation of Low-Capacity SG Blowdown system, until they are re-aligned to 1 -LW-TK-1 3.
- 3. Verify the following actions have occurred. IF the actions have NOT occurred, THEN ensure the following equipment alignment:
a) 1-LW-PCV-115-CLOSED b) 1-LW-FCV-100, Holdup Tank Influent Valve CLOSED c) SGBlowdownpumps-TRIPPED jS
- 4. Isolate Low-capacity SG Blowdown by closing the SG Blowdown HCVs on both Units.
- 5. Do the following to determine if the release has exceeded allowable limits:
a) Evaluate sample results and compare to the limits in the EAL Matrix Categories as applicable for plant conditions, to determine the need to implement EPIPs. b) IF it is NOT necessary to implement EPIPs, THEN initiate notifications specified in VPAP-2802, NOTIFICATIONS AND REPORTS.
NUMBER ATTACHMENT TITLE ATTACHMENT 0-AP-5.1 5 CLARIFIER OUTLET RADIATION MONITOR REVISION PAGE 28 3of3
- 6. Determine source of activity by use of the following as required:
a) Chemistry samples b) Health Physics analysis c) Liquid Waste release path valve verification d) Associated Radiation Monitor activity:
- Clarifier Inlet (1 -RM-LW-1 10)
- SG Blowdown (1 -SS-RM-1 22, 1-SS-RM-123, 1-SS-RM-124, 2-SS-RM-222, 2-SS-RM-223, 2-SS-RM-224, 1 -SS-RM-1 25, 2-SS-RM-225)
- Condenser Air Ejector (1-SV-RM-121, 2-SV-RM-221)
- Service Water discharge to Lake (1-RM-SW-108)
- Service Water Reservoir (1-RM-SW-109)
- CC HX Service Water (1-RM-SW-107)
- 7. HZ Unit 1 SG tube leakage is indicated, THEN initiate 1-AP-24, STEAM GENERATOR TUBE LEAK, or 1 -AP-24.1, SHUTDOWN STEAM GENERATOR TUBE LEAK, as appropriate, while continuing with this procedure.
- 8.
HZ Unit 2 SG tube leakage is indicated, THEN initiate 2-AP-24, STEAM GENERATOR TUBE LEAK, or 2-AP-24.1, SHUTDOWN STEAM GENERATOR TUBE LEAK, as appropriate, while continuing with this procedure.
- 9. Restore Liquid Waste and SG Blowdown release path(s) as directed by Health Physics.
- 10. RETURN TO 0-AP-5.1, COMMON UNIT RADIATION MONITORING SYSTEM, Step in effect.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 54. 073A2.02 054/BANKINAPS/H14/2.7/3.2/7/
Given the following conditions:
- A WGDT release is in progress.
- HI and HI-HI alarms are received on 1-GW-RM-178-1, Process Vent RM Noble Gas Normal.
- The cause of the alarm was determined to be the detector spiking.
- 1-GW-RM-178-1 indication is trending back to normal.
Which ONE of the following describes the automatic action that the OATC will verify, AND includes the procedural direction given to re-establish the WGDT release once the detector is operating normally? A. Verify 1-GW-FCV-1O1, WGDT to Process Vents, CLOSED; the WGDT release can be re-established once the OATC verifies the HI alarm clears. B. Verify 1-GW-FCV-1O1, WGDT to Process Vents, CLOSED; the WGDT release can be re-established once the Instrument Department resets the HI-HI alarm AND restores the MGP system to normal range monitoring. C. Verify 1-GW-TV-106, Equipment Vents, CLOSED; the WGDT release can be re-established once the OATC verifies the HI alarm clears. D. Verify 1-GW-TV-106, Equipment Vents, CLOSED; the WGDT release can be re-established once the Instrument Department resets the HI-HI alarm AND restores the MGP system to normal range monitoring.
- a. Incorrect. First part is correct; other valves are addressed by the procedure but this is the only one with an auto-close feature. Second part is incorrect but plausible since it would be true if the spike were only high enough to pick up the Hi alarm. Further if that candidate has the misconception that these auto actions are associated with the Hi alarm function only and not either/or they would select this distractor.
- b. Correct. First part is correct as noted above. Second part is also correct as described in O-AP-5.2, attachment 3. It should be noted that the Hi alarm must also be clear, thus the question contains the statement once the monitor is indicating normally, the candidate should conclude since the hi alarm will clear without any operator action required, that there is no Hi alarm present.
- c. Incorrect. First part is incorrect but plausible since as discussed in distractor a this valve is addressed by the procedure which gives an option of closing it, but again the stem specifically solicits knowledge of automatic actions. Second part also incorrect but plausible as discussed in Distractor a.
- d. Correct. First part is incorrect but plausible as discussed above. Second part is correct as explained in answer b.
QUESTIONS REPORT for NAPS 2010 RD NRC Exam rev3 Process Radiation Monitoring (PRM) System Ability to (a) predict the impacts of the following malfunctions or operations on the PRM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 /45.3 /45.13) Detector failure Tier: 2 Group: 1 Technical
Reference:
O-AP-5.2 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
NUMBER PROCEDURE TITLE REVISION 22 O-AP-5 2 MGP RADIATION MONITORING SYSTEM
. PAGE 3 of 3 H_STEP [H ACTION/EXPECTEDRESPONSE I RESPONSENOTOBTAINED I
- 1. INITIATE THE APPROPRIATE ATTACHMENT(S) FOR THE AFFECTED MGP MONITOR(S)
FOR ANY OF THE FOLLOWING CONDITIONS: (Continued) NIT RADIATION MONITOR RECORDER ATT. NO. MGP PROCESS VENT 1 -GW-RI-1 78-3 1-RM-RR-178 ATTACHMENT 3 1 -GW-RI-1 78-1 1 -G W-RI-1 78-2 MGP VENT STACK A 1 -VG-RI-1 79-3 1 -RM-RR-1 79 ATTACHMENT 4 1 -VG-RI-1 79-1 1 -VG-RI-1 79-2 MGP VENT STACK B 1 -VG-RI-1 80-3 1-RM-RR-180 ATTACHMENT 5 1 -VG-RI-1 80-1 1 -VG-RI-1 80-2
- 2. ALL MGP RADIATION MONITORS D RETURN TO Step 1.
THAT ARE SUSPECTED TO BE ABNORMAL HAVE BEEN CHECKED IN ACCORDANCE WITH THIS PROCEDURE
- 3. RETURN TO PROCEDURE AND STEP IN EFFECT
- END -
NUMBER ATTACHMENT TITLE ATTACHMENT O-AP-5.2 3 1-GW-RI-178-1, 2 OR 3, MGP PROCESS VENT RAD MONITOR REVISION PAGE 22 lof5 NOTE: The MGP Remote Display Unit indicating a fault will buzz with an audible alarm. Pressing the BLUE Select/Acknowledge key on the display unit will silence the audible alarm.
- 1. Check any MGP Process Vent Radiation Monitor(s) in alarm:
C
- l-GW-RI-178-3, Process Vent RM Particulate alert or high C
- 1 -G W-RI-1 78-1, Process Vent RM Noble Gas Normal - alert, high or H/H C
- 1-GW-RI-178-2, Process Vent RM Noble Gas Accident alert or high
- 2. IF at least one of the following conditions is met, THEN GO TO Step 4:
C
- 1-GW-RI-178-3, is in alarm and the Operate light is LIT.
C
- 1-GW-RI-178-1, is in alarm and the Operate light is LIT.
C
- 1-GW-Rl-178-2, is in alarm; and 1-GW-RI-178-3, is in standby and the Test light is LIT; and 1-GW-Rl-178-1, is in standby and the Test light is LIT.
- 3. IF a MGP Process Vent Radiation Monitor(s) Operate Light is NOT LIT, THEN GO TO Step 6.
- 4. Check Radiation Reading on alarming MGP Process Vent Radiation Monitor(s) as applicable:
a) Screen display using the channel keys:
- 1-GW-RI-178-3, Process Vent RM Particulate channels:
C
- A Part-Rel mci/sec C
- B Part-Act mci/cc
- 1-GW-Rl-178-1, Process Vent RM Noble Gas Normal channels:
C
- A NG-Rel-L mci/sec C
- B NG-Act-L mci/cc
- 1-GW-Rl-178-2, Process Vent RM Noble Gas Accident channels:
C
- A NG-Rel-H mci/sec C
- B NG-Act-H mci/cc C b) 1-RM-RR-178, Process Vent RM Activity and Release Rate recorder
NUMBER ATTACHMENT TITLE ATTACHMENT 0-AP-5.2 3 1-GW-Rl-178-1 2 OR 3, MGP PROCESS VENT RAD MONITOR REVISION PAGE 22 2of5
- 5. IF the abnormality of the Radiation Monitor was NOT caused by an obvious malfunction, THEN do the following for the MGP Process Vent Radiation Monitor(s) in alarm:
a) Inform the Health Physics Shift Supervisor of the following:
- Date and time the monitor alarmed and which channels are alarming
- Process Vent flow rate (1-GW-Fl-108)
NOTE: The analysis should be completed immediately due to time limits for event classification in EPIP-1 .01, EMERGENCY MANAGER CONTROLLING PROCEDURE. b) Request a sample and survey of the affected area. c) IF a Hi alarm has actuated on 1 -GW-RI-1 78-3 OR 1 -GW-RI-1 78-1, THEN verify the following:
- 1-GW-TV-102A CLOSED
- 1-GW-TV-102B CLOSED J
D
- 1-GW-FCV-101 - CLOSED t
- 1-CV-P-3A, Unit 1 Containment Vacuum Pump A STOPPED-
- 1-CV-P-3B, Unit 1 Containment Vacuum Pump B STOPPED-
- 2-CV-P-3A, Unit 2 Containment Vacuum Pump A STOPPED-
- 2-CV-P-3B, Unit 2 Containment Vacuum Pump B - STOPPED d) Ensure the following trip valves are closed, as directed by the SRO, to stop potential radioactive sources to the Process Vent System:
- 1-GW-TV-106, Equipment Vents
- 1-GW-TV-1 13, Liquid Waste Tank Vents
- 1-GW-TV-114, Boron Recovery Tank Vents e) Do the following to determine if the release has exceeded allowable limits:
- 1) Evaluate sample results, dose projections, and meter readings and compare to the limits in the EAL Matrix Categories as applicable for plant conditions, to determine the need to implement EPIPs.
- 2) IF EPIP implementation is NOT required, THEN initiate notifications specified in
( VPAP-2802, NOTIFICATIONS AND REPORT.
NUMBER ATTACHMENT TITLE ATTACHMENT 0-AP-5.2 3 1-GW-RI-178-1, 2 OR 3, MGP PROCESS VENT RAD MONITOR REVISION PAGE 22 3of5 NOTE:
- IF Hi alarm is lit on 1-GW-Rl-178-1 or 1-GW-RI-178-3, THEN WGDT discharge and D V 2
4 c*c Containment Vacuum pumps will be locked out until the Hi alarm(s) clear. r
- A Hi Hi (H/H) alarm on 1-GW-Rl-178-1 will swap monitoring to 1-GW-RI-178-2, Process Vent RM Noble Gas Accident. The WGDT discharge and Containment Vacuum pumps will be locked out until the Hi Hi latch is released by the Instrument Department. The MGP system is returned to normal range monitoring by the Instrument Department.
- IF power was lost to the MGP Radiation Monitor unit, THEN observe the following:
When power is restored, then the MGP system will startup in accident mode.
- After a loss of power, the MGP system should automatically return to operable normal range monitoring approximately 20 minutes following power restoration.
- A burned out operating light bulb at the MGP skid should not render the radiation monitor (1 inoperable. The MGP display will indicate the affected bulb.
- 6. Do the following for the affected MGP Process Vent Radiation Monitor(s):
a) Check for any Faults indicated at bottom of screen display. b) IF a GRN LIGHT fault is display, THEN change the light bulb at the skid. c) Have the Instrument Department check the MGP Radiation Monitor(s) for operability and repair as applicable. d) IF a FLOW FAULT is indicated on 1 -G W-RI-1 78-1, THEN have the Instrument Department inspect the filter paper for discoloration and boron particles using 0-GIP-9.0, MGP PARTICULATE FILTER ASSEMBLY REPLACEMENT. (STEP 6 CONTINUED ON NEXT PAGE)
NUM BER ATTACHMENT TITLE ATTACHMENT O-AP-5.2 3 1-GW-RI-178-1, 2CR 3, MGP PROCESS VENT RAD MONITOR REVISION PAGE 22 4of5 e) IF a Hi Hi (H/H) alarm has swapped monitoring to the 1-GW-RI-178-2, Process Vent RM Noble Gas Accident Rad Monitor, AND 1-GW-RI-178-1 and 1-GW-Rl-178-2 are reading normal, THEN with Health Physics concurrence, do the following:
- 1) Have the Instrument Department reset the Hi Hi alarm AND restore the MGP system to normal range monitoring.
- 2) Reset the latch in logic to reset the Hi Hi alarm:
- a. On the PCS homepage, select P&ID MENU.
- b. On the PIDMENU screen, select RM - Radiation Monitoring.
- c. On the 1 RM1 screen, select Reset Rad Monitor Latches.
- d. On the RMRESET screen, select RESET PROCESS VENT LATCH.
f) IF a MGP Process Vent Radiation Monitor(s) is inoperable OR will be inoperable due to maintenance, THEN GO TO Step 7. g) WHEN the MGP Process Vent Radiation Monitor(s) is determined to be operable, THEN GOTO Step 8.
- 7. Declare the MGP Process Vent Radiation Monitor(s) inoperable:
a) Notify the Health Physics Shift Supervisor of the date and time the monitor was declared inoperable. b) Enter the monitor into the Action Statement Status Log, TRM 3.3.7, Table 3.3.7-1, Function 2.c. c) Initiate required actions of VPAP-2103N, Offsite Dose Calculation Manual (North Anna). d) Enter a Condition Report, as applicable. e) Notify Health Physics to begin obtaining the required contingency samples. f) Return to Step in effect in this attachment.
- 8. WHEN the source has been isolated or repaired, THEN restore the release paths as directed by the SRO.
NUMBER ATTACHMENT TITLE ATTACHMENT 0-AP-5.2 3 1-GW-Rl-178-1, 2 OR 3, MGP PROCESS VENT RAD MONITOR REVISION PAGE 22 5of5
- 9. Verify MGP Process Vent Radiation Monitor(s) operable. IF NOT, THEN RETURN TO Step 1:
a) 1-GW-RI-178-3, Process Vent RM Particulate: D
- operate light-LIT D
- testlight-NOTLIT D
- alert and high lights NOT LIT D
- Display operable b) 1-GW-RI-178-1, Process Vent RM Noble Gas Normal:
- operate light LIT EJ
- testlight-NOTLIT
- alert, high and H/H lights NOT LIT
- Display operable c) 1-GW-RI-178-2, Process Vent RM Noble Gas Accident:
- operate light LIT D
- test light LIT
- alert and high lights - NOT LIT D
- Display operable d) Unit 2 Annunciator Panel B-C5, PROCESS VENT VNT STACK A&B RAD MONITORS FAILURE NOT LIT e) Check 1 -GW-RM-1 78-2, Process Vent Radiation Monitor in 307 Switchgear TIC alarm light -
NOT BLINKING.
- 10. Do the following:
a) Notify the Health Physics Shift Supervisor that the MGP Process Vent Radiation Monitor(s) are operable. b) IF required, THEN restore Containment Vacuum Pumps and WGDT release using applicable procedures. c) Return to procedure and step in effect.
- END -
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 55. 074EK1 .03 055/BANK/NAPS: 5995/L/3/4.5/4.9/4/
Per 1-FR-Cl, Response to Inadequate Core Cooling, SGs are depressurized at the maximum rate to (ultimately) atmospheric pressure in order to A. reduce RCS temperature to collapse steam voids in the reactor vessel head. B. reduce SG pressure to enable low-pressure sources of SG feedwater to restore RCS heat sink. C. reduce RCS temperature to increase thermal driving head for natural circulation. D. reduce RCS pressure to allow the SI accumulators and LHSI pumps to inject water to the RCS.
- a. Incorrect. Plausible since head voiding would be indicative of core uncovery; the candidate who lacks detailed knowledge of the bases may conclude that this strategy would be a logical approach to quenching the core and reducing temperatures.
- b. Incorrect. Plausible since loss of secondary heat sink is a potential precurser to loss of core cooling; and again the candidate without detailed knowledge of the procedure and background document may conclude that this cause-and-effect relationship is the procedural strategy for cooling the core.
- c. Incorrect. Plausible since the candidate may assume that safety sources for injecting water are unavailable and thus this strategy could be used to at least buy time while further recovery actions continue.
- d. Correct. As stated in FR-C.l Background Document, The rapid secondary depressurization has been shown to be the most effective way to reduce RCS pressure. RCS pressure must be reduced in order for the SI accumulators and low-head SI pumps to inject.
Inadequate Core Cooling Knowledge of the operational implications of the following concepts as they apply to the Inadequate Core Cooling (CFR 41.8 /41.10 /45.3) Processes for removing decay heat from the core Tier: 1 Group: 2 Technical
Reference:
FR-C.1 & WOG background document Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
NUMBER PROCEDURE TITLE REVISION 14 1 -FR-C 1 RESPONSE TO INADEQUATE CORE COOLING
. PAGE 12 of 26
-]_STEP j ACTION/EXPECTEDRESPONSE I I RESPONSENOTOBTAINED I CAUTION: If Steam Dumps are used for RCS cooldown, then, to prevent an undesired Main Steamline Isolation, each Main Steamline flow should be kept less than 1 .0E6 LBM/HR. NOTE: Partial uncovering of SG tubes is acceptable in the following steps. 13._ DEPRESSURIZEALLINTACTSGsTO / 190 PSIG: 7w9e--. LAA 5 D a) Dump steam to Condenser at maximum D a) Dump steam using SG PORVs. rate H SG PORVs are NOT available, THEN dump steam using any of the following: Decay Heat Release Valve:
- a. Locally open isolation valve(s) for NON-RUPTURED SG(s) to Decay Heat Release Valve:
EJ
- 1 -MS-i 9, A Steam Line to i-MS-HCV-104 Non-Return Valve
- 1 -MS-58, B Steam Line to 1-MS-HCV-104 Non-Return Valve
- 1 -MS-96, C Steam Line to 1-MS-HCV-104 Non-Return Valve (STEP 13 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 14 1 -FR-C 1 RESPONSE TO INADEQUATE CORE COOLING
. PAGE 16 of 26
-_STEP ACTION/EXPECTEDRESPONSE RESPONSENOTOBTAINED I
- 17. DEPRESSURIZE ALL INTACT SGs TO D Dump steam using SG PORVs.
ATMOSPHERIC PRESSURE BY DUMPING STEAM TO CONDENSER AT MAXIMUM SG PORVs are NOT available, THEN RATE dump steam using any of the following: S
- Decay Heat Release Valve:
mV 1) Locally open isolation valve(s) for NON-RUPTURED SG(s) to Decay Heat Release Valve: D
- 1-MS-19, A Steam Line to 1-MS-HCV-104 Non-Return Valve D
- 1-MS-58, B Steam Line to 1-MS-HCV-104 Non-Return Valve
- 1-MS-96, C Steam Line to 1-MS-HCV-104 Non-Return Valve
- 2) Locally open 1 -MS-20, Decay Heat Release Valve Upstream Isolation Valve.
D 3) Manually open Decay Heat Release Valve. D
- Local operation of the SG PORVs
STEP DESCRIPTION TABLE FOR FR-C.1 Step 11 STEP: Depressurize All Intact SGs To (0.07) PSIG Joçs 4m (< PURPOSE: To recover the core via SI accumulator injection BASIS: The rapid secondary depressurization has been shown to be the most effective way to reduce RCS pressure. RCS pressure must be reduced in order for the SI accumulators and low-head SI pumps to inject. To prevent accumulator nitrogen injection, the operator should stop the secondary depressurization when the SG pressure reaches (0.07) psig and when VCIJSK at least two RCS hot leg temperatures fall below (F.05)°F. A steam generator 4 r pressure limit is set to preclude significant nitrogen injection into the RCS. To determine the steam generator pressure limit, an ideal gas expansion calculation should be performed based on nominal plant specific values for initial accumulator tank pressure (P ), initial nitrogen gas vol ume (V 1 ), and 1 final nitrogen gas volume (V ). The final nitrogen gas volume should be 2 equivalent to the total accumulator tank volume. The RCS pressure at empty tank conditions (P ) is determined from: 2 Y V 1 P = v V 2 P where y = 1.25 for ideal gas expansion. The steam generator pressure limit is then determined by subtracting the RCS to SG delta P from P . The RCS to 2 SG delta P should be calculated as described in the RCP TRIP/RESTART section in the Generic Issues of the Executive Volume. Instrument uncertainties are not included in the determination of the steam generator pressure limit to preclude a bias toward either having more accumulator water injected into the RCS or having less nitrogen injected into the RCS. The hot leg temperature of (F.05)°F should be determined so that the RCS saturation pressure exceeds the accumulator pressure after the accumulator water has been discharged. This precludes nitrogen injection into the RCS. To determine the hot leg temperature, an ideal gas expansion calculation should be performed based on nominal plant specific values for initial accumulator tank pressure (P ) , initial nitrogen gas volume (V 1 ) , and final 1 nitrogen gas volume (V ). The final nitrogen gas volume should be equivalent 2 to the total accumulator tank volume. The RCS pressure at empty tank conditions (P ) is determined from: 2 V 1 P = Y V 2 P Where y = 1.25 for ideal gas expansion. The setpoint temperature of (F.05)°F FR-C.1 Background 36 HP-Rev. 2, 4/30/2005 HFRC1BG .doc
STEP DESCRIPTION TABLE FOR FR-C.1 Step 11 is the saturation temperature corresponding to P . Instrument uncertainties 2 are not included in the determination of the RCS hot leg temperature setpoint to preclude a bias toward either having more accumulator water injected into the RCS or having less nitrogen injected into the RCS. ACTIONS: o Determine if SG pressures are less than (0.07) psig o Determine if SG pressure is decreasing o Determine if at least two RCS hot leg temperatures are less than (F.O5)°F o Determine if RCS hot leg temperatures are decreasing o Dump steam to condenser at maximum rate o Dump steam at maximum rate using SG PORVs o Stop SG depressurization INSTRUMENTATION: o SG pressure indication o RCS hot leg temperature indication o Position indication for:
- Condenser steam dump valves - SG PORVs CONTROL/EQUIPMENT:
o Switches for:
- Condenser steam dump valves - SG PORVs KNOWLEDGE:
N/A FR-C.1 Background 37 HP-Rev. 2, 4/30/2005 HFRC1BG.doc
STEP DESCRIPTION TABLE FOR FR-C.l Step 11 PLANT-SPECIFIC INFORMATION: o Plants that have qualified RCS wide range pressure transmitters may use RCS pressure instead of RCS hot leg temperature to determine when the accumulator should be isolated. o Plants that have qualified pressure transmitters on the accumulators may use accumulator pressure instead of RCS hot leg temperature to determine when the accumulators should be isolated. o (0.07) Minimum SG pressure which prevents accumulator nitrogen injection. Refer to background document for FR-C.l. o (F.O5) RCS hot leg temperature to prevent accumulator nitrogen injection. Refer to background document for FR-C.l. FR-C.1 Background 38 HP-Rev. 2, 4/30/2005 HFRC1BG .doc
STEP DESCRIPTION TABLE FOR FR-C.1 Step 14 STEP: Depressurize All Intact SGs To Atmospheric Pressure AS PURPOSE: To recover the core via low-head safety injection S BASIS: With continued SG depressurization, RCS pressure should follow secondary pressure until the shutoff head of the low-head SI pumps is reached. Then, low-head safety injection should begin to refill the RCS. V U ACTIONS: SDC O o Depressurize all intact SGs to atmospheric pressure by dumping steam to condenser at maximum rate o Depressurize all intact SGs to atmospheric pressure by dumping steam to condenser at maximum rate using SG PORVs INSTRUMENTATION: o SG pressure indication o Position indication for:
- SG PORVs - Condenser steam dump valves CONTROL/EQUIPMENT:
Switches for: o SG PORVs o Condenser steam dump valves KNOWLEDGE: N/A PLANT-SPECIFIC INFORMATION: N/A FR-C.1 Background 43 HP-Rev. 2, 4/30/2005 HFRC1BG .doc
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 56. 076K! .07 056/NEW//L/3/2.5/2.3/4/
Which ONE of the following identifies the normal and backup cooling water sources to the Control Room Chillers? A. Bearing Cooling normally supplies both Units Control Room Chillers; either Unit can be supplied from Service Water as a backup. B. Bearing Cooling normally supplies both Units Control Room Chillers; ONLY Unit 1 can be supplied from Service Water as a backup. C. Service Water normally supplies both Units Control Room Chillers; either Unit can be supplied from Bearing Cooling as a backup. D. Service Water normally supplies both Units Control Room Chillers; ONLY Unit 1 can be supplied from Bearing Cooling as a backup.
- a. Incorrect. Plausible since both are a source and the candidate may confuse which is normal and which is backup.
- b. Incorrect. Plausible as noted above.
- c. Incorrect. First part is correct. Second part incorrect but plausible, if the candidate is unaware of this Unit difference they will select this distractor.
- d. Correct. Service Water is the normal source and only Unit 1 has the capability of being supplied from Bearing Cooling.
Service Water System (SWS) Knowledge of the physical connections and/or cause- effect relationships between the SWS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8) Secondary closed cooling water Tier: 2 Group: 1 Technical
Reference:
1-OP-21.11 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info: This is a Unit differences question.
Procedure: I -OP-21 .11 Rev: 002 PAR: 0 T[tle: BEARING COOLING SUPPLY TO THE UNIT I CONTROL ROOM CHILLERS Effective Date: 07/15/2003 Station: North Anna CONTINUOUS USE
DOMINION 1-OP-21.11 North Anna Power Station Revision 2 Page 4 of 18 2.3.6 12050-FM-80A, Unit 2 Bearing Cooling Water System 2.3.7 Safety Evaluation 92-SE-OT-067, Rev. 2. 2.4 Commitment Documents None 3.0 INITIAL CONDITIONS 3.1 The Unit 1 Bearing Cooling Water System is in service. 3.2 Review the equipment status to verify station configuration will support performance of this procedure. 3.3 One of the following conditions exists: A loss of the Service Water headers to the Unit 1 Control Room Chillers has occurred. OR
- Prior NRC approval has been obtained to use Bearing Cooling to supply the Unit 1 Control Room Chillers during maintenance or outage.
4.0 PRECAUTIONS AND LIMITATIONS 4.1 Comply with the following guidelines when marking steps N/A:
- IF the conditional requirements of a step do not require the action to be performed, THEN mark the step N/A.
- IF any other step is marked N/A, THEN have the Shift Supervisor (or designee) approve the N/A and justify the N/A on the Procedure Routing Sheet.
4.2 Tech Spec 3.7.11 requires two MCR/ESGR ACS subsystems to be operable in Modes 1, 2, 3, and 4, and during movement of recently irradiated fuel assemblies.
DOMINION 1-OP-21.11 North Anna Power Station Revision 2 Page 5 of 18 4.3 Use of this procedure during maintenance, outages, or situations other than emergencies is not allowed without prior NRC approval. 4.4 Supplying Unit 1 Control Room Chillers with Bearing Cooling makes them inoperable in Modes 1, 2, 3 and 4. 4.5 Use good Station House Keeping practices by routing vents and drains to floor drains before opening the vent / drain valve.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 57. 077AA1.02 057/NEW//L/3/3.8/3.7/6/
Which ONE of the following parameters can be used to discriminate between a voltage regulator malfunction and a grid disturbance? A. Generator Megawatts. B. Generator Megavars. C. Generator Output Voltage. D. Exciter Field Current.
- a. Correct. Swing in megawatts are indicative of grid instability, and while they can also be caused by irregularities in the turbine control system, the stem clearly addresses comparing the grid and voltage regulator.
- b. Incorrect. Incorrect but plausible, as vars can be affected by either, but the candidate who does not understand operation of the voltage regulator may erroneously assume that changing vars means there is an issue with the voltage regulator, while this is true, a grid disturbance can also affect vars.
- c. Incorrect. Incorrect but plausible, as voltage can be affected by either, but the candidate who does not understand operation of the voltage regulator may erroneously assume that changing voltage means there is an issue with the voltage regulator, while this is true, a grid disturbance can also affect voltage.
- d. Correct. Incorrect but plausible, as this can be affected by either, but the candidate who does not understand operation of the voltage regulator may erroneously assume that changing field current is indicative of an issue with the voltage regulator, while this is true, a grid disturbance can also affect exciter field current.
Generator Voltage and Electric Grid Disturbances Ability to operate and/or monitor the following as they apply to Generator Voltage and Electric Grid Disturbances: (CFR: 41.5 and 41.10 /45.5, 45.7, and 45.8) Turbine / generator controls Tier: 1 Group: 1 Technical
Reference:
0-AP-8, 1-AP-26 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
NUMBER PROCEDURE TITLE REVISION 6 O-AP-8 RESPONSE TO GRID INSTABILITY PAGE 4 of 7 H_STEP ACTION! EXPECTED RESPONSE j RESPONSE NOT OBTAINED 7._ CHECK BOTH UNITS STATION SERVICE D Initiate 0-AP-lO, Loss Of Electrical Power, BUSES ENERGIZED as required. NOTE: Fluctuations in mega-watts (spikes and dips) or mega-watt swings indicate grid disturbance. Voltage Regulator failures have little to no affect on mega-watt output. The Voltage Regulators should be kept in AUTO during grid disturbances. 8._ VERIFY BOTH OPERATING UNITS D available, THEN place Voltage VOLTAGE REGULATORS IN AUTO - Regulators in AUTO. 9._ CHECK IF LOW VOLTAGE SYSTEM 11 GOTO Step 11. CONDITIONS EXIST: D
- 500 KV voltage drops below 505 KV OR D
- Notification from System Operator that an Actual LOW EMERGENCY LIMIT has been triggered OR D
- Contingency Analysis indicates that an Emergency Limit will be exceeded, if a trip of a North Anna Unit was to occur 10._ DECLARE OFFSITE POWER INOPERABLE AND ENTER APPLICABLE ACTIONS:
D
- Tech Spec 3.8.1, AC Sources Operating Modes 1, 2, 3, and 4 D
- Tech Spec 3.8.2, AC Sources Shutdown Modes 5 and 6
NUMBER PROCEDURE TITLE REVISION 9 1 -AP-26 FAILURE OF MAIN GENERATOR VOLTAGE REGULATOR HIGH PAGE 2 of 3 H_STEP ACTION/EXPECTEDRESPONSE I RESPONSENOTOBTAINED CAUTION: IF unsure of conditions, THEN leave the Voltage Regulator in Auto. 1] CHECK GENERATOR STATUS: Do the following: D
- Sustained Generator output l
- Contact System Operator voltage increase EJ
- Return to procedure and step in effect AND D
- MVAR sustained increase consistent with Generator j output voltage AND C
- MW stable 2] TURN VOLTAGE REGULATOR CONTROL SWITCH TO OFF 3 ] ADJUST GENERATOR VOLTAGE Do the following:
USING THE VOLTAGE REGULATOR BASE ADJUST C a) HD Reactor power is greater than or equal to SWITCH: 30%, THEN GO TO 1 -E-0, REACTOR TRIP OR SAFETY INJECTION. C
- GENERATOR OUTPUT VOLTAGE - NORMAL RANGE C b) H Reactor power is less than 30%, THEN Trip Turbine and GO TO 1 -AP-2.1, TURBINE TRIP C
- MVARs NOT greater than WITHOUT REACTOR TRIP REQUIRED.
200 MVARs IN
- 4. NOTIFY SYSTEM OPERATOR THAT VOLTAGE REGULATOR IS IN BASE CONTROL
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 58. 078K2.O1 058/BANK//L/2/2.7/2.9/8/
Which ONE of the following identifies the power supplies to the Unit 1 and Unit 2 Instrument Air compressors, 1-IA-C-i and 2-IA-C-i, respectively? 1-IA-C-i is powered from and 2-IA-C-i is powered from A. iH Emergency Bus; 2H Emergency Bus B. iH Emergency Bus ; 2J Emergency Bus C. iJ Emergency Bus; 2H Emergency Bus D. iJ Emergency Bus; 2J Emergency Bus
- a. Correct. The H train on each Unit supplies that Units Instrument Air Compressor.
- b. Incorrect. Plausible since the candidate who lacks detailed systems knowledge might assume it would be logical to have the compressors powered from opposite trains.
- c. Incorrect. Plausible as discussed in Distractor b.
- d. Incorrect. Again the candidate who lacks detailed systems knowledge, might know that the IA compressors come from Emergency Busses, and the SA Compressors come from station service busses, but may not know which Emergency bus for which IA compressor.
Instrument Air System (lAS) Knowledge of bus power supplies to the following: (CFR: 41.7) Instrument air compressor Tier: 2 Group: 1 Technical
Reference:
1-OP-26A & 2-OP-26A Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
DOMINION 1 -0P-26A North Anna Power Station Revision 50 Page 55 of 141 (Page 5 of 5) Attachment 19 I -EE-MCC-1 HI -2S, I HI -2S Motor Control Center I -EP-MC-20 r - LOCATION: CABLE VAULT POWER SUPPLY: 1 -EE-BKR- 1 4H 1-7
*v 7ccusm
REFERENCE:
1 1715-FE-1Q j - Breaker Required md No. Load Position Verifier Verifier 1 -HV-F-40A, M4 On Safeguards Area Exhaust Fan Circuit Breaker 1-EP-CB-41AN, Heat Trace Distribution Panel Circuit Cold Weather On Breaker, AND N2L l-EP-CB-41BN Heat Trace Distribution Circuit Breaker Hot Weather Off Hot Weather 1 -EP-CB-14N, N2R On Heat Trace Distribution Panel Ckt Bkr 1 -CH-P- 1 Cl, N3 On 1C Charging Pump Aux Oil Pump Circuit Breaker 1 -SW-P-5, N4 On Rad Monitor Sample Pump Circuit Breaker P1 1-IA-C-i AUX BLDG INSTR AIR COMP On MODE 1-4 Off REC-29, 480 Volt Power Receptacle ** P2L MODE 5 &6 On Circuit Breaker 1 -IC-DRIV- 1 D, P2R On Incore Instrumentation Drive Assembly Ckt Bkr l-CH-P-1A1, P3 On Aux Oil Pump For 1-CH-P-1A Ckt Bkr P4 On 1-RM-P-159A & B Sample Pump Feed
- IF breaker is on with the Unit in Mode 5 or 6, THEN have the SRO enter into the Action Statement Status Log to open the Breaker prior to entering Mode 4. (Reference 2.1.2)
DOMINION 2-OP-26A North Anna Power Station Revision 47 Page 52 of 163 (Page 3 of 4) Attachment 19 2-EE-MCC-2H1 -2S LOCATION: CABLE VAULT POWER SUPPLY: 2-EE-BKR-24H-3
REFERENCE:
12050-FE-i N Breaker Required md No. Load Position Verifier Verifier 2-CH-MOV-23 80, Hi On Seal Water Return Inside Isol Circuit Breaker 2-RC-MOV-2595, H2 Locked Off C Reactor Cool Loop Cold Leg Isol Valve Ckt Bkr 2-RC-MOV-25 92, H3 Locked Off B Reactor Cool Loop Hot Leg Isol Valve Ckt Bkr 2-RS-MOV-20 1 B, H4 On B Casing Cooling Pump Discharge Isol Valve CB 2-RC-MOV-25 93, Ji Locked Off B Reactor Cool Loop Cold Leg Isol Valve Ckt Bkr 2-HC-HC- 1, J2L On Unit 2 Hydrogen Recombiner Circuit Bkr i-EP-CB-129, J2R On Alt Feed To 1 -HC-H2A- 101 Ckt Bkr J3L Spare, J3L Off Spare Circuit Breaker 2-EP-CB-84A1, J3R On Quench Spray Area Motor Heater Cabinet Ckt Bkr 1 -HV-F-75B, On Auxiliary Building Appendix R Fan Ckt Bkr 2-IA-C-i, Ki On Instrument Air Cprsr 1 Circuit Breaker 2-IC-DRIV-1D, K2L On Incore Instrumentation Drive Assembly Ckt Bkr K2R Off Spare 2-DA-P-iA, K3 On 1 A Safeguards Area Sump Pump Circuit Breaker
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 59. 086K5 .03 059/NEW//H/3/3 .1/3.4/8/
If a Main Transformer deluge is inadvertently actuated, the Electric Fire Pump is started by and the associated Main Transformer will A. a limit switch on the deluge valve; de-energize B. a limit switch on the deluge valve; remain energized C. a pressure switch on the fire main ; de-energize D. a pressure switch on the fire main; remain energized
- a. Incorrect. First part is plausible since several systems (including fire protection) have automatic features associated with valve position logic and since this is a large load on the FP system the candidate who lacks detailed systems knowledge would rationalize that this feature makes sense from a design standpoint. Second part is incorrect but plausible since again the candidate who lacks detailed systems knowledge will choose this because they will feel it is the more conservative system response.
- b. Incorrect. First part is plausible as discussed above. Second part is correct while certain conditions that would cause deluge actuation will lockout the transformer (e.g. sudden fault pressure), the actuation of the deluge itself however will not.
- c. Incorrect. First part is correct a pressure switch is the automatic feature that will start the subject pump when the demand (from the deluge operation) exceeds the ability of the jockey pumps to maintain system pressure. Second part incorrect but plausible as discussed in distractor a.
- d. Correct. First part is correct as discussed in distractor c. Second part also correct as discussed in distractor b.
Fire Protection System (FPS) Knowledge of the operational implication of the following concepts as they apply to the Fire Protection System: (CFR: 41.5 /45.7) Effect of water spray on electrical components Tier: 2 Group: 2 Technical
Reference:
ARs D-C6, D-F6, dwgs 11715-ESK-5BB & 11715-LSK-1-3D Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
VIRGINIA POWER 1EI-CB-21D ANNUNCIATOR C6 1-AR-D-C6 NORTH ANNA POWER STATION REV. 0 SNSOC APPROVAL: ON FILE Effective Date:06/29/95 MOTOR FIRE PP RUNNING 1.0 Probable Cause 1.1 Selector switch in AUTO and fire main pressure < 90 psig 1.2 START pushbutton depressed 1.3 Selector switch in ON position 2.0 Operator Action 2.1 IF actual fire, THEN GO TO 0-FCA-0, Fire Protection - Operations Response. 2.2 Determine fire pump started on low Fire Main pressure. 2.3 Verify if a test is in progress. 3.0 References 3.1 11715ESK5BA, 5BB 3.2 11715FEiC, 1J 3.3 0-FCA0, Fire Protection Operations Response 4.0 Actuation 4.1 43MDF, Selector switch 4.2 1PB, START pushbutton 4.3 1FPPS1202
-END-
VIRGINIA POWER l-EI-CB-21D ANNUNCIATOR F6 1-AR-D-F6 NORTH ANNA POWER STATION REV. 0 SNSOC APPROVAL: ON FILE Effective Date:06/29/95 FIRE WATER SYSTEM INITIATED 1.0 Probable Cause 1.1 Any Unit 1 Deluge System heat detector above setpoint 1.2 Spray pushbutton on Fire Protection Panel 1.3 Manual Release pushbutton on Deluge Panel 1.4 Any local 3way test switch in Test position 1.5 Operation of any Unit 1 local Break Glass Station 1.6 Flow to Reactor Containment Standpipe 1.7 Flow to Auxiliary Building sprinklers 1.8 Manual Release at any Manual Fire Station Operator Action 2.1 Refer to 1AR-19, Fire Protection Panel. 2.2 IF actual fire, THEN GO TO 0FCA-0, Fire Protection Operations Response. 3.0 References 3.1 1AR19, Fire Protection Panel 3.2 OFCA-0, Fire Protection - Operations Response 3.3 11715LSK201A, 201B, 20ic, 20iD 3.4 ii7i5ESK1OD, 1OBAH 3.5 Instrument Loops i17l5FP080, 081, 083, and 084 4.0 Actuation 4.1 Any heat detector above setpoint 4.2 Flow in any Deluge System loop header downstream of isolation valve 4.3 Water flow to either Unit 1 or Unit 2 Reactor Containments 4.4 Spray pushbutton depressed at Fire Protection Panel for any Zone 4.5 Manual Release pushbutton depressed at Deluge Panel for any Zone 4.6 Any 3-way switch in Test position 4.7 Water flow to Auxiliary Building sprinklers 4.8 Any Break Glass Station Release 4.9 Any Manual Fire Station Release
-END-
1 NUCLEAR ENGINEERING RICHMOND.VIRGINIA REVISED PER SEP 5-PI/XR 2øE-526. DRAWING ND THIS DWG SUPERSEDES REV 6. 1 1715LSK 13D PCEC1 & MCI 13-JUN-20g6 ØV:2V PRIOR TO USING FOR DESIGN WORK CHECK OMIS FOR WORK PENDING 1enn31
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1- > 2ST MOT 0 UMPER I NO. 3 (PE) KR 3 a (PF) PVR) RUN TrN TART FAIL Ii RE TELUE VALI E I rUN REL\Y RELAY I/SLAY FIELA/ Li 2 Fl, TIMER 2 jeif5E PF2. FAILU. I 2U TPIX 57 R LA-ç NOTE5 f. A,H ¶; L_ : E MOTOR 7RVEJ4 PIRE PUMP CKT FPPNO5 NoTE Z. PREFIX M EXTERNP\L WIRIN5 REFER TO MFR DWG 5&W FILE NO. i.IO3 B FPPNG5 ALL MF/? SUPPLIED RELMS ARE TYPE &E5 cc t- Pc c IT! S oorc T!fCTS ALL EQUIPMEM9 AT FIRE PUMP 5iI2 IAl fubf-;t_L S INT&k a IP1E COMTROLLER. UNLESS oTNERJlSE ELEF-iTN I-:-: 4icWCKi5 IJOICATEb. MOTOR RS: Fi1I I-Lt9 H) NORTH AN}I- Dv.RPc -TiflQN 2-Po5. LPRli-ft 1-2.ET TO AJTO MAD REV IRIN(A CC CTZII CP- opEIA-roc SrOlcx & WEBSTER ENGINEERING COR?OR.TION OT-F- O-I (-T BLOK I115ESK-5BB A 52:54 I
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 60. 103K3.03 060/NEW//H/4/3.7/4.1/5/
The crew is preparing for Unit 1 core off-load. Which ONE of the following describes the method specified in 1-PT-91, Containment Penetrations, to control containment penetrations, AND includes the approval required to breach a penetration during fuel movement? Containment penetrations shall be tagged with A. Danger tags, and an SRO or Shift Manager may approve breaching a penetration. B. Danger tags, and FSRC approval is required for breaching a penetration. C. Caution/Information tags, and an SRO or Shift Manager may approve breaching a penetration. D. Caution/Information tags, and FSRC approval is required for breaching a penetration.
- a. Incorrect. Plausible since this would indicate the most positive administrative control and was in fact the method used in the past, so the candidate who lacks knowledge of the subject or relies on past knowledge will likely select this distractor. Second part is correct, safety analysis supports this so the SM or SRO is an appropriate level of authority to control this.
- b. Incorrect. First part incorrect but plausible as noted above. Second part is incorrect but plausible; given the sensitivity of the issue this is again a case where the candidate who lacks detailed knowledge may select this because it would indicate a higher level of authority, and thus greater scrutiny. Since FSRC approves most administrative controls it would make sense to the person who lacks understanding that they would also have to be the ones to relax or take exception to them.
- c. Correct. First part is correct; as discussed above this is a change to past methods. Second part correct as discussed in Distractor a.
- d. Incorrect. First part is correct as discussed in Distractor c. Second part is incorrect but plausible as discussed in distractor b.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Containment System Knowledge of the effect that a loss or malfunction of the containment system will have on the following: (CFR: 41.7 I 45.6) Loss of containment integrity under refueling operations. Tier: 2 Group: 1 Technical
Reference:
1 -PT-9 I 1-LOG-i 8 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
DOMINION 1-PT-91 North Anna Power Station Revision 35 Page 8 of 99 4.8 Instructions, such as the following statement, should be placed on ALL Caution/Information Tags prior to hanging the tag(s) in the field: Maintenance is NOT allowed on this component. This component supports containment integrity. Do NOT operate without Shift Manager permission. Contact Operations at extension 2541 for further guidance. 4.9 Attachment 7, Human Factors To Consider When Clearing PT-91 Tags, provides Human Factors guidance when clearing 1-PT-91 tags. 4.10 Tagouts written for this procedure should reference the use of Attachment 7 when clearing tags. 5.0 SPECIAL TOOLS AND EQUIPMENT None I I I t ( .
DOMINION 1-PT-91 North Anna Power Station Revision 35 Page 14 of 99 NOTE: The approvals of Substep 6.1 .7.d are not required to NOT hang Caution/Information Tags on the items listed in Step 6.1.10 that have no means of access.
- d. IF not desired to hang Caution/Information Tags on all valves which will be closed, THEN obtain the following approval signatures, dates, and times for approvals. (Reference 2.4.3)
I Manager Nuclear Operations Date/Time Manager Nuclear Outage and Planning Date/Time (C 1
/
tjJ SNSOC Date/Time N 6 1 8 Complete Attachment 1 as follows
- a. Close the remote manual valves listed on Attachment 1 for systems NOT in service, as indicated by the SROs initials, and initial on the attachment in the appropriate spaces.
- b. IF desired, THEN mark, with identifying tape, the controls for any remote operated valves that are used as a boundary.
- c. IF any remotely operated valve does NOT have position indication in the Control Room (such as valve de-energized), THEN hang a Caution/Information Tag on the valve.
6.1.9 Complete Attachment 2 and Attachment 3 as follows:
- a. Determine if tags will be hung on isolation valves (tags required unless approval in Step 6.1.7 has been obtained).
DOMINION 1-PT-91 North Anna Power Station Revision 35 Page 18 of 99 6.2 Establishing Containment Boundaries Before Movement Of Irradiated Fuel Assemblies Within Containment 6.2.1 Record the following:
- a. Date Subsection 6.2 is initiated:
- b. Time Subsection 6.2 is initiated:
NOTE: Recently Irradiated Fuel is defined in the Bases of Tech Spec 3.9.4, Containment Penetrations, as fuel that has occupied part of a critical reactor core within the previous 100 hours. The following step verifies that Tech Spec 3.9.4 is not applicable. 6.2.2 Verify that the reactor has been shut down for greater than 100 hours and therefore no fuel meets the criteria for Recently In-adiated Fuel. NOTE: The following step ensures compliance with the 7-day reverification of Containment isolation in accordance with Ops management requirements. 6.2.3 Do the following to ensure documentation of containment closure capability is maintained:
- a. Enter into the Action Statement Status Log and initiate a 1-LOG-i 4 to complete l-PT-91 reverification (Subsection 6.3) once every 5 days from the date and time recorded in Step 6.2.1. (This surveillance may be suspended when movement of irradiated fuel assemblies within Containment is complete.)
- b. IF the reverification interval will exceed the 5 day requirement, THEN obtain the Manager Nuclear Operations permission to exceed the 5-day requirement.
DOMINION 1-PT-91 North Anna Power Station Revision 35 Page 19 of 99 6.2.4 IE the Equipment Hatch is desired to be open during movement of irradiated fuel in containment, THEN do the following:
- a. Ensure that the following requirements of the NRC Safety Evaluation for Tech Spec Change N-Oil for Alternate Source Term are satisfied:
(Reference 2.4.9)
- 1. Ensure that 1-LOG-i 8, Containment Boundary Breach Log, is being maintained to track containment openings. (Reference 2.4.9)
- 2. Ensure that pre-designated individuals, including radiological protection personnel, have been trained and briefed and will be available to perform O-MCM-1204-05, Emergency Installation of the Equipment Door and Temporary Penetration Plate (When RCS Level is More Than 42 Inches Above Centerline). (Reference 2.4.9)
- 3. Ensure that cables and hoses that penetrate the Equipment Hatch are provided with quick disconnects. (Reference 2.4.9)
- 4. Ensure that tools and equipment needed to perform Equipment Hatch closure have been pre-staged. (Reference 2.4.9)
- b. Do the following to ensure that the hatch can be closed:
- 1. Verify that the Equipment Hatch Temporary Hatch Plate is installed.
- 2. Verify that the Equipment Hatch is entered into i-LOG-i 8, Containment Boundary Breach Log.
6.2.5 the Equipment Hatch is desired to remain closed during movement of irradiated fuel in containment, THEN do the following:
- a. Verify that the Equipment Hatch is closed and is held in place by at least four bolts.
DOMINION 1-PT-91 North Anna Power Station Revision 35 Page 20 of 99
- b. Verify integrity of the Equipment Hatch Escape Lock opening by verifying that one of the following conditions is met:
- The Equipment Hatch Escape Lock is installed and at least one air lock door is closed.
- The Equipment Hatch Temporary Hatch Plate is installed.
CAUTION All mechanical penetrations through the equipment hatch must have check valves installed to prevent air out-leakage from Containment or be controlled by another method to prevent air out-leakage. (References 2.4.2 and 2.4.4) 6.2.6 the Equipment Hatch Temporary Hatch Plate is installed, THEN do the following:
- a. Ensure that the status of the Equipment Hatch Temporary Hatch Plate is maintained as follows:
- 1. Have the Containment Coordinator initiate 0-MCM-1202-03, Control C Coord .
of Equipment Hatch Temporary Hatch Plate Ports and Penetrations. (References 2.4.2, 2.4.4, and 2.4.5)
- 2. Inform the Containment Coordinator to ensure that 0-MCM-1202-03 SRO is accurately maintained during the movement of irradiated fuel assemblies within Containment. (Reference 2.4.5)
DOMINION 1-PT-91 North Anna Power Station Revision 35 Page 21 of99
- b. Enter in LCO Tracking and initiate a 1-LOG-14 to perform the following once every 24 hours during movement of irradiated fuel assemblies within Containment:
- 1. Verify the following attachments of 0-MCM- 1202-03, Control of Equipment Hatch Temporary Hatch Plate Ports and Penetrations, are up to date: (References 2.4.2, 2.4.4, and 2.4.5)
- Attachment 4, Line Installation Data Sheet SRO
- Attachment 7, Line Removal Data Sheet SRO
- 2. Verify Attachment 6, Equipment Hatch Temporary Hatch Plate SRO Isolation Verification Log, of 0-MCM- 1202-03, Control of Equipment Hatch Temporary Hatch Plate Ports and Penetrations, has been completed for the day. (References 2.4.2, 2.4.4, and 2.4.5) 6.2.7 Verify at least one Personnel Hatch airlock door is capable of being closed as follows: (Reference 2.4.8)
- a. Complete 1-OP-18.1A, Valve CheckoffContainment Access.
- b. Verify that no cables or hoses are run through the airlock.
- c. Verify that any seating surface protective devices (such as protective coverings, equipment ramps, and so on) are capable of being removed in a timely manner.
- d. IF both Personnel Airlock doors are to be left open, THEN do the following:
- 1. Close and reopen the Personnel Airlock doors using 1-OP-i 8.1 to
( demonstrate operability.
DOMINION 1-PT-91 North Anna Power Station Revision 35 Page 22 of 99
- 2. Verify that the Personnel Airlock is entered into 1-LOG-i 8, Containment Boundary Breach Log.
6.2.8 Verify that both Personnel Air Lock Emergency Escape Locks are closed with the operating handles are in the closed position and Cautionllnformation Tags installed on each handle (this requires four Caution/Information Tags). (Reference 2.4.7) 6.2.9 Verify that all electrical penetrations are installed in the following areas: (Reference 2.4.7)
- Cable Vault Elect Elect
- Fuel Building Elect Elect 6.2.10 IF the transfer canal is NOT flooded, THEN verify that the Fuel Transfer Tube Gate Valve is closed. (Reference 2.4.7) 6.2.11 Verify one of the following conditions exist for RVLIS Train A and B:
- a. RVLIS Channels A and B are operating as expected by ALL of the following being meet:
- The expected reading is indicated.
- Channel check is satisfactory.
- Hydraulic Isolator is NOT in alarm on diagnostic page.
- b. IF RVLIS condition CANNOT be verified, THEN have an Instrument Inst Inst IV . .
Technician verify RVLIS capillary tubing is filled with PG water OR verify that the RVLIS capillary tubing outside of containment, its hydraulic isolator, and associated fittings are intact.
DOMINION 1-PT-91 North Anna Power Station Revision 35 Page 23 of 99 6.2.12 Do the following to ensure that containment purge isolation is functional:
- a. Record initial status of Containment Purge Supply and Exhaust Fans by placing a check in the box for each running fan:
1-HV-F.-4A, CONT PURGE SUPPLY FAN 1-HV-F-4B, CONT PURGE SUPPLY FAN 1-HV-F-5A, CONT PURGE EXH FAN 1-HV-F-5B, CONT PURGE EXH FAN
- b. IF Unit 2 Containment Purge is in progress, THEN secure Unit 2 Purge and close Unit 2 Purge Valves during this test in accordance with 2-OP-21 .2, Containment Purge.
- c. Place the control switches for the following fans in STOP:
- 1-HV-F-4A, CONT PURGE SUPPLY FAN
- 1-HV-F-4B, CONT PURGE SUPPLY FAN
- 1-HV-F-5A, CONT PURGE EXH FAN
- l-HV-F-5B, CONT PURGE EXH FAN
DOMINION 1-PT-91 North Anna Power Station Revision 35 Page 24 of 99
- d. Open the following MOVs:
- l-HV-MOV-100A, Cont Purge Supply Vlv
- l-HV-MOV-100B, Cont Purge Supply Vlv
- 1 -HV-MOV-i OOC, Cont Purge Exh Vlv
- 1 -HV-MOV-i OOD, Cont Purge Exh Vlv
- 1 -HV-MOV- 101, Cont Purge Exh Bypass Vlv
- 1-HV-MOV- 102, Cont Purge ATM Relief Valve
- e. Close the following MOVs using the control switches on the ventilation panel:
- 1 -HV-MOV-i OOA, Cont Purge Supply Vlv
- 1-HV-MOV-100B, Cont Purge Supply Vlv
- 1 -HV-MOV-i OOC, Cont Purge Exh Vlv
- 1-HV-MOV-100D, Cont Purge Exh Vlv
- 1-HV-MOV-10i, Cont Purge Exh Bypass Vlv
- 1-HV-MOV-102, Cont Purge ATM Relief Valve
DOMINION 1-PT-91 North Anna Power Station Revision 35 Page 25 of 99
- f. IF desired, AND Unit 1 Containment Purge valves are OPERABLE OR Unit 1 is in Mode 5, THEN return Unit 1 Containment Purge and Exhaust System to service in accordance with l-OP-21.2, Containment Purge. Refer to Step 6.2.12.a for initial fan status.
- g. IF desired AND Unit 2 Containment Purge valves are OPERABLE OR Unit 2 is in Mode 5, THEN return Unit 2 Containment Purge and Exhaust System to service in accordance with 2-OP-2 1.2, Containment Purge. Refer to Step 6.2. 12.a for initial fan status.
6.2.13 Have the SRO do the following: NOTE: Some valves may be isolated outside of Containment while other valves on a different system may be isolated inside of Containment.
- a. Determine if valves will be closed inside of Containment or outside of SRO Contamment.
- b. Initial in the appropriate spaces on the attachments to designate whether penetrations are isolated inside of Containment or outside of Containment.
- c. Initiate 1-LOG- 19, Penetration Isolation Status Log.
SRO 6.2.14 Complete Attachment 1 as follows:
- a. Close the remote manual valves listed on Attachment 1 for systems NOT in service, as indicated by the SROs initials, and initial on the attachment in the appropriate spaces.
- b. IF desired, THEN mark, with identifying tape, the controls for any remote operated valves that are used as a boundary.
- c. IF any remotely operated valve does NOT have position indication in the Control Room (such as valve de-energized), THEN hang a Caution/Information Tag on the valve.
DOMINION 1-PT-91 North Anna Power Station Revision 35 Page 26 of 99 6.2.15 Complete Attachment 2 and Attachment 3 as follows:
- a. Close the manual valves listed on Attachment 2 and Attachment 3 for systems NOT in service, as indicated by the SROs initials, and initial on the attachments in the appropriate spaces. Document on Attachment 2 and Attachment 3 as required. (Reference 2.4.1)
- b. IF the penetration boundaries for the Steam Generators or RSHX are to be inside containment, THEN check required handholes, manways and flanges installed and intact.
- c. Hang required tags. IF Check valves, Safety valves, or Relief valves are used as boundaries, THEN also place Tags on those valves.
(Reference 2.4.3)
DOMINION 1-PT-91 North Anna Power Station Revision 35 Page 27 of 99 NOTE: The following step determines if boundaries are accessible by ladders or scaffolding and directs Caution/Information Tags be placed on those that are accessible to ensure the boundaries will not be opened. IF no access is available, THEN the port and valve is assumed to be closed, with the exception of 1-WT-39, l-WT-51, and 1-WT-67 which are assumed to be open. 6.2.16 IF the penetration boundaries for the Steam Generators will be inside Containment, THEN place a Caution/Information Tag on any item below that has a ladder, scaffolding, or other means of access erected to it. N/A the items that have no access:
- l-WT-39, A Steam Generator Chemical Feed Downstream Isolation
- l-FW-256, 1A Steam Gen Feedwater Inlet Header Drain Valve
- 1 -WT-5 1, B Steam Generator Chemical Feed Downstream Isolation
- l-FW-258, lB Steam Gen Feedwater Inlet Header Drain Valve
- l-WT-67, C Steam Generator Chemical Feed Downstream Isolation
- 1 -FW-260, 1 C Steam Gen Feedwater Inlet Header Drain Valve
- 1-MS-3, 1A S/G Main Steam Line Vent
- l-RC-E-1A Inspection #1, l-RC-E-1A Inspection Port
- l-RC-E-1A Inspection #2, l-RC-E-1A Inspection Port Step continued on the next page
DOMINION 1-PT-91 North Anna Power Station Revision 35 Page 28 of 99 Step continued from the previous page:
- 1-MS-42, lB S/G Main Steam Line Vent
- 1-RC-E-1B Inspection #1, l-RC-E-1B Inspection Port
- l-RC-E-1B Inspection #2, 1-RC-E-1B Inspection Port
- l-MS-80, 1C S/G Main Steam Line Vent
- 1-RC-E-1C Inspection #1, 1-RC-E-1C Inspection Port
- l-RC-E-lC Inspection #2, l-RC-E-1C Inspection Port
( 6.2.17 GO TO Section 7.0.
DOMINION 1-LOG-18 North Anna Power Station Revision 5 Page 6 of 10 5.0 INSTRUCTIONS 5.1 The containment boundary breach log will be established (at the direction of the SRO or the OMOC) prior to RCS level being lowered to 42 inches above centerline. 5.2 The containment boundary breach log will be established (at the direction of the SRO or the OMOC) prior to and during movement of irradiated fuel in Containment. 5.3 The containment boundary breach log will be established (at the direction of another procedure, the SRO, or the OMOC) whenever containment vacuum is broken. 5.4 The SRO shall maintain control of the containment boundary breach log. 5.5 The SRO shall authorize all containment boundary beaches and log all appropriate information on Attachment 1, Containment Boundary Breaching Log V 1 5.6 During operation of the Reactor Coolant System at decreased inventory, ALL containment boundary breaches must be able to be sealed by the containment closure team within < 1 hour. (Reference 2.3.1) 5.7 During operation of the Reactor Coolant System at reduced inventory, ALL containment boundary breeches must be able to be sealed by the Containment Closure Team within less than the time to core boiling, as determined by 1-GOP-i 3.0, Alternate Core Cooling Method Assessment, and 1-GOP-i 3.1, Alternate Core Cooling Method Assessment Guidelines. (Reference 2.3.2) 5.8 During movement of irradiated fuel in Containment, to maintain the concept of defense-in-depth and assuming acceptable radiological protection conditions exist after a fuel handling accident in containment, containment closure will be established within 45 minutes following the decision to isolate containment. (Reference 2.4.2)
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 61. G2.1.19 061/BANK!IL/2/3.9/3.8//
In accordance with l-OP-2.l, Unit Startup from Mode 2 to Mode 1, during a Unit ramp, all of the following indications should be used to monitor real power EXCEPT A. U980, Power, Calorimetric B. U 1201, Power, 1 Mm Avg STM C. U1203, Power, 1 Mm Avg FW D. U1231, Power, 1 Mm Avg UFM
- a. Correct. While this is the primary calorimetric power (NIs are compared to and adjusted to this value during performance of periodic tests to satisfy Tech Spec surveillance requirement 3.3.1.2 for calorimetric heat balance calculation) during steady state operation, as stated in 1-OP-2.1 it should not be relied upon during a ramp since it will lag real power.
- b. Incorrect. This is an alternate power indication provided by PCS that the operator should use during a ramp as stated in 1-OP-2.1. The candidate who lacks detailed understanding of the limitations of the primary calorimetric point (U0980) would likely eliminate this choice since by virtue of the title 1 mm avg, it implies a delay which is not implicit in the title of the choice that is the answer.
- c. Incorrect. Incorrect but plausible as discussed above.
- d. Correct. Incorrect but plausible as discussed above.
Conduct of Operations Ability to use plant computers to evaluate system or component status. CFR: 41.10 /45.12) Tier: 3 Technical
Reference:
1-OP-2.1 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
PROCEDURE NO: Dornnon 1-OP-2.1 REVISION NO: NORTH ANNA POWER STATION 95 PROCEDURE TYPE: UNIT NO: OPERATING PROCEDURE PROCEDURE TITLE: UNIT STARTUP FROM MODE 2 TO MODE I SURV ICCE REACT REQ III MGT REVISION
SUMMARY
Made the following changes for DCP 07-02 1, Feedwater UFM Installation PCS -
- Revised synopsis statement calorimetric power (Ui 203) to calorimetric power (U 1231, Power, 1 Mm Avg UFM or U 1203, Power, 1 Mm Avg FW). Added a table to the synopsis to show power level indications.
- Added U1231, Power, 1 Mm Avg UFM to P&L 4.25, Caution for Step 5.2.1 and Caution for Step 5.2.79.
In Step 5.2.9(c) Changed Operations Manager on Call to Senior Operations Manager because OMOC is not required for ICCE III per OP-AA-106. In Step 5.2.56 added cross reference to Step 5.2.55 to identify which annunciators. PROBLEMS ENCOUNTERED: EJ NO YES Note: If YES, note problems in remarks. REMARKS: (Use back for additional remarks.) SRO: DATE: CONTINUOUS USE
DOMINION 1-OP-2.1 North Anna Power Station Revision 95 Page 9of98 Uncertainties in the calculated calorimetric power level increase when reactor power level is below 100% RTP. This increased uncertainty can be introduced into the Power Range NI channels when they are normalized to the calorimetric power at these power levels. This can potentially cause the NI High Flux Trip Setpoints to be non-conservative if the NI Channels are adjusted downward at these lower power levels. To prevent this from happening, the NI High Flux Trip Setpoints and NI Rod Stop Setpoints may need to be reduced if the NIs are adjusted downward at reduced calculated calorimetric power levels. This ensures that the effective high flux trip setpoints remain below 118% RTP safety analysis value by accommodating potentially increased calorimetric errors at reduced power. No adjustment to the overpower trip and rod stop setpoint will be required if the NIs are not adjusted downward. (References 2.3.39 and 2.4.16) An important aspect of the Feedwater based calorimetric is the response time of indicated power (U0980, Power, Calorimetric). Since the normal feedwater flow oscillations are excessive, the feedwater based calorimetric uses software filters to smooth out the oscillations in the flow inputs. However, the filters also increase the response time to power changes, such that the indicated power will lag real power. The lag for step changes in power is much greater than that for ramp changes, and the lag time and time to settle out are directly related to the magnitude of the step change or ramp rate. For example, during a ramp of 0.3% per minute starting at 90% power, indicated power will lag real power by approximately 1.5% after about 15 minutes. If the ramp were stopped at 98.5% indicated power, actual power would be about 100%. Indicated power would reach 100% power in about 20 minutes. Conversely, if the ramp is stopped at 100% indicated power, real power would be approximately 101.5%. Therefore the 1-minute unfiltered average FW calorimetric power (U123 1, Power, 1 Mm Avg UFM or U1203, Power, 1 Mm Avg FW) should be frequently compared to U0980 to avoid overshooting 100% power. (Reference 2.3.43) PCS Point:
Description:
U1231 Power, 1 Mm Avg UFM U 1203 Power, 1 Mm Avg FW
DOMINION 1-OP-2.1 North Anna Power Station Revision 95 Page 21 of98 4.22 Uncertainties in the calculated calorimetric power level increase when reactor power level is below 100% RTP. This increased uncertainty can be introduced into the Power Range NI channels when they are normalized to the calorimetric power at these power levels. This can potentially cause the NI High Flux Trip Setpoints to be non-conservative if the NI Channels are adjusted downward at these lower power levels. To prevent this from happening, the NI High Flux Trip Setpoints and NI Rod Stop Setpoints must be reduced if the NIs are adjusted downward at reduced calculated calorimetric power levels. This ensures that the effective high flux trip setpoints remain below 118% RTP safety analysis value by accommodating potentially increased calorimetric errors at reduced power. No adjustment to the overpower trip and rod stop setpoint will be required if the NIs are not adjusted downward. (References 2.3.39 and 2.4.16) 4.23 For the initial startup following refueling, stabilize Reactor power at greater than 28 percent but NOT to exceed 30 percent, as indicated by Calorimetric for the first flux map. 4.24 IF a Reactor trip occurs after the second Feedwater Pump is started (45 percent power) AND before the second unisolated Feedwater recirc is placed in AUTO, THEN the second unisolated Recirc should be placed in AUTO to ensure adequate Feedwater recirc with two Feedwater Pumps since Feedwater demand is reduced. (Reference 2.3.40) 4.25 During a transient and when approaching a upper or lower power limit do NOT rely on U0980, Power, Calorimetric, to represent real power. U0980 lags real power due to its filtered flow input. Power indications U123 1, Power, 1 Mm Avg UFM, U1203, Power, 1 Mm Avg FW, and U1201, Power, 1 Mm Avg STM, will more accurately represent real power. (Reference 2.3.43) 4.26 IF PMT is required on ANY MFRV, THEN ensure 1-PT-2 13.24, Valve Inservice Inspection (Feedwater Regulator Valves Partial Stroke), is also performed during this startup. (Reference 2.3.46)
DOMINION 1-OP-2.1 North Anna Power Station Revision 95 Page 25 of 98 5.2 Unit Startup from Mode 2 to Mode 1 CAUTION Do NOT rely on U0980, Power, Calorimetric, to represent real power during a Unit ramp. U0980 lags real power due to its filtered flow input. Use alternate power indications U1231, Power, 1 Mm Avg UFM, U1203, Power, 1 Mm Avg FW, and U1201, Power, 1 Mm Avg STM, to monitor real power. NIs, ATs, SG pressure, and First Stage pressure may also be used to monitor real power. During steady state conditions, U0980 is the primary calorimetric power. (Reference 2.3.43) 5.2.1 Verify Initial Condition is satisfied. 5.2.2 Review Precautions and Limitations. 5.2.3 Review the active Temporary Modifications to ensure that the active SRO . Temporary Modifications will not adversely affect the unit startup from Mode 2 to Mode 1. (Reference 2.4.22) 5.2.4 the Unit is already on line and this procedure is being entered to raise power, THEN make the following notifications to Energy Supply (MOC): (Reference 2.3.52)
- Notify Energy Supply (MOC) at least 30 minutes prior to commencing power increase.
- Notify Energy Supply (MOC) when power increase is initiated.
5.2.5 IF blowdown of condenser hotwell is in progress and the Operations Manager On Call (OMOC) has NOT given permission to continue hotwell blowdown, THEN to prevent an inadvertent reactivity addition due to increased feed flow to the Steam Generators from an inadvertent Condensate pump start, secure blowdown of condenser hotwell. (Reference 2.4.17)
DOMINION 1-OP-2.1 North Anna Power Station Revision 95 Page 62 of 98 CAUTION
- Alternate power indications (including NIs, z\Ts, SG pressure, First Stage pressure, MWe, and 1 Mm Avg Feed Flow / 1 Mm Avg Steam Flow Calorimetrics), MUST be monitored to ensure excessive disparity does NOT exist between parameters. The SRO should be notified of any excessive disparity.
- Approach 100% Reactor power slowly and do NOT rely on U0980, Power, Calorimetric, to represent real power. U0980 lags real power due to its filtered flow input. Use power indications U1231, Power, 1 Mm Avg UFM, U1203, Power, 1 Mm Avg FW, and U1201, Power, 1 Mm Avg STM, to monitor real power.
Typically for a 0.3% turbine step increase, allow approximately 15 minutes for U0980 to catch-up to real power. (Reference 2.3.43) NOTE: As Reactor power is increased to obtain 100% as indicated by U0980, Steam Flow based calorimetric points U1200 and U1201 may indicate greater than 100%. NOTE: Changing AFD and Xenon concentration may cause NI power to increase even though calorimetric power is stable. NIs may require adjustment to maintain NIs within Tech Spec tolerance. (Reference 2.4.24) NOTE: NI adjustments should be done in accordance with 1-PT-24.1, Calorimetric Heat Balance (Computer Calculation) (Reference 2.4.24) NOTE: Main Turbine power should be increased in small increments, with enough time allotted between the power increases for U0980 to catch-up to real power. NOTE: 100% Reactor power should be approached slowly and in such a manner as to compensate for reactivity effects and prevent exceeding 100% power. 5.2.79 WHEN Reactor power is 99% or less, THEN do the following:
- a. Stabilize Reactor power at 99% power.
- b. Hold power stable at 99% power for approximately 15 minutes.
(
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 62. G2.1.20 062/NEW//H14/4.6/4.6//
With Unit 1 at 100% power, the following conditions exist:
- 1-CW-MOV-1O1D, D Waterbox Inlet MDV, indicates mid-position.
- Condenser pressure is 2.2 inches Hg abs. and degrading.
Which ONE of the following identifies the immediate action, AND includes the action required by 1-AP-14, Low Condenser Vacuum, to mitigate this event? A. Verify at least two Circulating Water pumps running; Ensure 1-VP-i 8, Air Ejector suction valve for D water box, is CLOSED B. Verify condenser pressure: 3.5 inches Hg abs or less and stable; Ensure 1-VP-i 8, Air Ejector suction valve for D water box, is CLOSED C. Verify at least two Circulating Water pumps running; Ensure 1-CW-MOV-1 COD, D CW pump Discharge MDV, is OPEN D. Verify condenser pressure: 3.5 inches Hg abs or less stable; Ensure i-CW-MOV-100D, D CW pump Discharge MOV, is OPEN
- a. Correct. D CW pump is interlocked with this valve and will trip requiring implementing 1-AP-13; the given action is the IOA of that procedure. Second part is also correct; since there is no CW flow in the affected waterbox the air ejector suction must be closed to prevent steam binding.
- b. Incorrect. First part incorrect but plausible since the candidate who is unaware of the associated CW pump interlock or does not have complete knowledge of all procedural lOAs would likely select this distractor (since it is logical and is in fact a subsequent procedural action). The second part is correct as noted above.
- c. Incorrect. First part is correct as noted above. Second part is incorrect but plausible since it is an action required by 1-AP-14, however it is only performed for RUNNING CW pumps; as noted above D CW pump will trip on interlock, but the candidate who is not knowledgeable of the interlock feature may default to this distractor since as noted it would be correct if the pump were running.
- d. Incorrect. First part is incorrect but plausible as discussed in Distractor B. Second part also incorrect but plausible as discussed in Distractor C.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Conduct of Operations Ability to interpret and execute procedure steps. (CFR: 41.10 /43.5 /45.12) Tier: 3 Technical
Reference:
1-AP-13, 1-AP-14 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
NUMBER PROCEDURE TITLE REVISION 15 1 -AP-1 3 LOSS OF ONE OR MORE CIRCULATING WATER PUMPS PAGE 2 of 8 C j ACTION! EXPECTED RESPONSE RESPONSE NOT OBTAINED c (U v(tQ I C C- o -vr VERIFY AT LEAST TWO C GO TO 1-E-O, REACTOR TRIP OR WATER CIRCULATING PUMPS - SAFETY INJECTION, while continuing with this RUNNING procedure. i-i_ r
> , _T 1
- 2 VERIFY CONDENSER VACUUM: C Initiate 1-AP-14, LOW CONDENSER VACUUM, 1 while continuing with this procedure.
EJ
- 3.5 INCHES HG ABS OR LESS AND is (s ohs C
- STABLE I
- 3. CHECK CIRCULATING WATER Manually or locally close affected MOV:
PUMP DISCHARGE MOV ON ANY NON-RUNNING PUMP CLOSED-C
- 1-CW-MOV-100A C
- 1-CW-MOV-100B C
- 1-CW-MOV-100C C
- 1-CW-MOV-100D
- 4. CHECK ANY CIRCULATING WATER C GOTOStep5.
PUMP DEGRADED AS INDICATED BY LOW AMPS C a) Place affected Circulating Water Pump control switch in PTL C b) Verify affected pump discharge b) Manually or locally close affected MOV: MOV closed C
- 1-CW-MOV-100A C
- 1-CW-MOV-100B C
- 1-CW-MOV-100C C
- 1-CW-MOV-100D
rN Doii ien NORTH ANNA POWER STATION ABNORMAL PROCEDURE NUMBER PROCEDURE TITLE REVISION 23 1 -AP-1 4 LOW CONDENSER VACUUM (WITH FOUR ATTACHMENTS) PAGE 1 of 10 PURPOSE To provide instructions to follow in the event of a partial loss of Condenser vacuum. ENTRY CONDITIONS This procedure is entered when any of the following conditions exist:
- Degrading Condenser vacuum, or
- Decreasing Generator output, or
- i-AS-MOV-100, Condenser Vacuum Breaker, is OPEN, or
- Annunciator Panel G F-3, TURBINE LO VACUUM PRE-TRIP, is LIT, or
- Annunciator Panel G E-1, LP TURB 1A EXH HOOD HI TEMP, is LIT, or
- Annunciator Panel G F-i, LP TURB 1A EXH HOOD HI-HI TEMP, is LIT, or
- Annunciator Panel G G-i, LP TURB lB EXH HOOD HI TEMP, is UT, or
- Annunciator Panel G H-i, LP TURB 1 B EXH HOOD HI-HI TEMP, is LIT.
CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 23 1 -AP-14 LOW CONDENSER VACUUM PAGE 2 of 10 H_STEP ACTION/EXPECTED RESPONSE J RESPONSE NOTOBTAINED NOTE:
- IF Reactor power is 30% or greater AND Condenser pressure is greater than 5.5 inches Hg abs, THEN 1-E-0, REACTOR TRIP OR SAFETY INJECTION, should be entered.
- IF Reactor power is less than 30% AND Condenser pressure is greater than 3.5 inches Hg abs, THEN 1-E-0, REACTOR TRIP OR SAFETY INJECTION, should be entered.
- 1. INITIATE RCS BORATION USING THE FOLLOWING AS REQUIRED:
D
- ATTACHMENT 4, ESTIMATING BORATION FLOW RATES OR EJ
- Standard Ramp Plan NOTE: Ramp rates close to 5%/minute may cause the Steam Dumps to arm.
*2 REDUCE PLANT LOAD AT 5%! MINUTE OR LESS UNTIL CONDENSER VACUUM IS STABLE, AS FOLLOWS:
D a) Verify Turbine Load Control in a) IF Turbine is in OPERATOR AUTO, THEN do the IMP-IN following: D 1) Ensure Turbine control Reference and Setter values are matched. D 2) Place Turbine Load Control in IMP-IN by depressing the IMP-IN pushbutton. D b) Initiate turbine load reduction using ATTACHMENT 3, MAIN TURBINE OPERATION GUIDANCE in - OPERATOR AUTO or TURBINE MANUAL
NUMBER PROCEDURE TITLE REVISION 23 1 -AP-1 4 LOW CONDENSER VACUUM PAGE 3 of 10 J_STEP J ACTION/EXPECTEDRESPONSE I RESPONSENOTOBTAINED
- 3. VERIFY RODS IN AUTO D IF rod control is operable, THEN place rods in AUTO.
- 4. ENERGIZE ADDITIONAL PRZR HEATERS TO MAINTAIN PRZR PRESSURE ABOVE 2205 PSIG
*5 VERIFY PROPER AUTO Place Control Rods in MANUAL and adjust as CONTROL ROD INSERTION required to maintain Tavg within 5 °F of Tref. *6. MONITOR CONDENSER D H Reactor power is less than 30%, THEN GO TO PRESSURE 3.5 INCHES HG 1-E-0, REACTOR TRIP OR SAFETY INJECTION, ABS OR LESS while continuing with this procedure.
D IF Reactor power is 30% or greater AND Condenser pressure is greater than 5.5 inches hg abs, THEN GO TO 1-E-0, REACTOR TRIP OR SAFETY INJECTION, while continuing with this procedure.
- 7. SEND AN OPERATOR TO LOCALLY PERFORM ATTACHMENT 2, TURBINE BUILDING CORRECTIVE ACTIONS 8... VERIFY 1-AS-MOV-100, Manually close 1-AS-MOV-100, Condenser CONDENSER VACUUM BREAKER - Vacuum Breaker.
CLOSED
NUMBER PROCEDURE TITLE REVISION 23 1 -AP-14 LOW CONDENSER VACUUM PAGE 4 of 10 H_STEP ACTION/EXPECTEDRESPONSE I I RESPONSENOTOBTAINED 9._ CHECK CIRCULATING WATER SYSTEM STATUS: D a) INLET and OUTLET Water Box a) Do the following: MOVs FULL OPEN Ensure Air Ejector suction valve for any out of service water box is closed: D
- A Water box: 1 -VP-3 D
- BWaterbox:1-VP-4
- CWaterbox:1-VP-17
- DWater box: 1-VP-18 L.
EJ
- Initiate the following applicable procedure, as required while continuing with this procedure:
- 1-OP-48.2, Operation of Circulating Water System
- A Water Box 1 -MOP-48.30
- B Water Box 1 -MOP-48.31
- C Water Box 1-MOP-48.32
- D Water Box 1-MOP-48.33 (E b) Running Circulating Water Pumps Open affected Circulating Water Pumps discharge MOVs 100 PERCENT
- discharge MOVs while continuing with this
-OREJL. procedure:
- 1 -CW-MOV-1 QUA D
- 1-CW-MOV-100B D_1-CWMOV1OOC D
- 1-CW-MOV-100D Jz
QUESTIONS REPORT for NAPS 2010 RD NRC Exam rev3
- 63. G2.1.9 063/BANK!/L/3/2.9/4.5//
The Backboards operator is conducting an infrequently performed PT. One of the steps of the procedure is designated IV (Independent Verification). The Backboards operator feels that CV (Concurrent Verification) would be a more appropriate method of verification for the step. As the OATC, you review the procedure and concur with the Backboards operators assessment. Which ONE of the following describes the guidance provided by Pl-AA-500, Verification Practices, to address this issue? A. The OATC may authorize the use of CV in lieu of IV. B. A procedure change is the ONLY acceptable method of changing the type of verification. C. The Operations Manager MUST grant approval anytime CV is used in lieu of IV. D. The Shift Manager or SRO may authorize the use of CV in lieu of IV.
- a. Incorrect. Plausible since the candidate who is not familiar with PI-AA-500 would likely conclude that this method would be acceptable since the RO would control numerous PTs normally, and the check from Backboards to the higher line of authority would suffice.
- b. Incorrect. Plausible because a procedure change is of course an acceptable method, but lAW PI-AA-500 it is not the ONLY method. However because of the serious nature of verification practices the candidate may erroneously conclude that since this is a non-emergency situation that there is no other alternative.
- c. Incorrect. Plausible again because of the serious nature the candidate may conclude that a higher level of authority would be required.
- d. Correct. As stated in Pl-AA-500 this decision falls under SM/SRO authority.
Conduct of Operations Ability to direct personnel activities inside the control room. (CFR: 41.10 / 45.5 / 45.12/45.13) Tier: 3 Technical
Reference:
Pl-AA-500 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
Nuclear Fleet r J Dominion Administrative Procedure
Title:
Verification Practices Procedure Number Revision Number Effective Date and PI-AA-500 1 Approvals On File Revision Summary
- Revised 3.25 as follows:
- OLD INFORM the responsible supervisor upon completion of the task or NOTIFY the assigned controller.
- NEW INFORM the responsible supervisor upon completion of the task or NOTIFY the assigned verifier.
Functional Area Manager: Manager Nuclear Operations INFORMATION USE
DOMINION PI-AA-500 REVISION 1 PAGE 3 OF 14 1.0 PURPOSE NOTE: Concurrent Verification within the Dominion fleet is also referred to as Dual Verification (DV) or Simultaneous Verification (SV). Within this procedure, all references to Concurrent Verification shall also apply to DV and SV. Continued use of DV and SV in Dominion procedures is acceptable; however, these terms should be phased out and replaced as documents are revised for other purposes. This procedure establishes the fleet policy for component position verification. This includes Independent Verification (IV) and Concurrent Verification (CV). The intent of this procedure is to prevent the inadvertent mis-positioning of station equipment that could result in an unsafe condition. 2.0 SCOPE 2.1 This procedure applies to all personnel involved in positioning of station equipment and applies to all component position verification practices. Component position verification practices are used to confirm the correct configuration or status of equipment. 2.2 This procedure does not apply to peer checking practices. Peer checking is a human performance tool used to reduce human error. Human performance tools are described in Pl-AA-5000, Human Performance (HU). 3.0 INSTRUCTIONS 3.1 General NOTE: Attachment 1 contains general expectations for use of verification practices. 3.1.1 Where required, controlling documents shall specify the verification type to be used. NOTE: The Shift Manager or SRO may specify use of CV in lieu of IV. When this is necessary, consideration should be given to submitting a revision request for the controlling document. SRO 3.1.2 DETERMINE if CV is to be used in lieu of IV. 3.1.3 jf CV is to be used, THEN GO TO Subsection 3.3, Concurrent Verification. 3.2 Independent Verification (IV) Performer 3.2.1 SELF CHECK the correct component to be manipulated. 3.2.2 PERFORM the action specified in the controlling document. INFORMATION USE
DOMINION PI-AA-500 REVISION 1 PAGE 5 OF 14 Performer 3.3.6 EXECUTE the correct action on the correct component as directed by the controlling document. Verifier 3.3.7 jf the performers action is inconsistent with the controlling document, THEN DIRECT the performer to STOP the action. Performer 3.3.8 if the desired action was not performed correctly, THEN PLACE the equipment in a safe condition and NOTIFY the responsible supervisor. NOTE: Acceptable verification methods and At-Risk Practices to avoid are described in Attachment 1 General Expectations for Verification Practices. Performer and 3.3.9 if the desired action is performed successfully, THEN: Verifier
- a. Separately CONFIRM that the condition and the expected response are correct by one or more of the following means:
- 1. Hands-on check (preferred).
- 2. Remote indication:
- If multiple remote indicators are available, use as many as possible.
- If possible, perform at least one check locally to confirm remote indication.
- 3. System response.
- b. SIGN or INITIAL the controlling document to record the verification.
4.0 RECORDS None 5.0 ADMINISTRATIVE INFORMATION u cc i
ç C 5.1 Commitments None 5.2 Responsibilities 5.2.1 Shift Manager or SRO The Shift Manager or SRO is responsible for authorizing the use of CV in lieu of IV where not already specified by a controlling procedure.
5.2.2 Responsible Supervisor The Responsible Supervisor is responsible for authorizing the waiver of verification in accordance with the requirements of this procedure. INFORMATION USE
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 64. G2.2.2 064/NEW!/L/3/4.6/4.1//
Unit 1 was initially at 100% power. 1-AP-2.2, Fast Load Reduction, has been entered. Which ONE of the following is the correct sequence of actions as required by 1-AP-2.2? A. 1) Bypass AMSAC
- 2) Transfer Rod Control to MANUAL
- 3) Transfer Feedwater Control to Bypass FCVs B. 1) Bypass AMSAC
- 2) Transfer Feedwater Control to Bypass FCVs
- 3) Transfer Rod Control to MANUAL C. 1) Transfer Feedwater Control to Bypass FCVs
- 2) Bypass AMSAC
- 3) Transfer Rod Control to MANUAL D. 1) Transfer Rod Control to MANUAL
- 2) Bypass AM SAC
- 3) Transfer Feedwater Control to Bypass FCVs
- a. Incorrect. Plausible since there would be no severe consequence if done in a different order, given that the candidate who lacks detailed understanding may select any of the choices. Again there is no specific safety bases for the order, however the order is chosen to maximize availability and in some cases control of systems, thus minimizing the probability of an undesired plant transient.
- b. Correct. This is the order specified in 1-AP-2.2.
- c. Incorrect. Plausible since there would be no severe consequence if done in a different order, given that, the candidate who lacks detailed understanding, may select any of the choices. Again there is no specific safety bases for the order, however the order is chosen to maximize availability and in some cases control of systems, thus minimizing the probability of an undesired plant transient.
- d. Incorrect. Plausible since there would be no severe consequence if done in a different order, given that, the candidate who lacks detailed understanding, may select any of the choices. Again there is no specific safety bases for the order, however the order is chosen to maximize availability and in some cases control of systems, thus minimizing the probability of an undesired plant transient.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Equipment Control Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels. (CFR: 41.6 / 41.7 /45.2) Tier: 3 Technical
Reference:
1-AP-2.2 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
nen, NORTH ANNA POWER STATION ABNORMAL PROCEDURE NUMBER PROCEDURE TITLE REVISION 19 1-AP-2.2 FAST LOAD REDUCTION (WITH FIVE ATTACHMENTS) PAGE 1 of 18 PURPOSE To provide instructions in the event that plant conditions require a continuous load reduction at a sustained Turbine ramp rate from 1%/minute through 5%/minute. ENTRY CONDITIONS This procedure is entered when a fast load reduction (a sustained Turbine ramp rate from 1%/minute through 5%/minute) is required as determined by the SRO. CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 19 1-AP-2.2 FAST LOAD REDUCTION PAGE 2 of 18 H_STEP_fI ACTION/EXPECTEDRESPONSE I RESPONSENOTOBTAINED I NOTE: Boration is required during the ramp to maintain Control Rods above the Lo-Lo Insertion Limits. 1._ INITIATE RCS BORATION USING THE FOLLOWING AS REQUIRED:
- ATTACHMENT 5, ESTIMATING BORATION FLOWRATES OR
- Standard Ramp Plan NOTE: Ramp rates close to 5%/minute may cause the Steam Dumps to arm.
2._ REDUCE PLANT LOAD AT 5%! MINUTE OR LESS AS FOLLOWS: D a) Verify Turbine Load Control in a) IFTurbineisinOPERATORAUTO,THENdothe IMP-IN following: D 1) Ensure Turbine control Reference and Setter values are matched. D 2) Place Turbine Load Control in IMP-IN by depressing the IMP-IN pushbutton. E b) Initiate turbine load reduction using ATTACHMENT 4, MAIN TURBINE OPERATION GUIDANCE in - OPERATOR AUTO or TURBINE MANUAL
NUMBER PROCEDURE TITLE REVISION 19 1-AP-2.2 FAST LOAD REDUCTION PAGE 3 of 18 -_STEP ACTION! EXPECTED RESPONSE H RESPONSE NOT OBTAINED 3._ VERIFY RODS IN AUTO C IF rod control is operable, THEN place rods in AUTO. 4._ ENERGIZE ADDITIONAL PRZR HEATERS TO MAINTAIN PRZR PRESSURE ABOVE 2205 PSIG
*5 VERIFY PROPER AUTO C Place Control Rods in MANUAL and adjust as CONTROL ROD INSERTION required to maintain Tavg within 5 O F of Tref.
6._ MONITOR STEAM DUMPS FOR PROPER OPERATION
*7 MAINTAIN THE FOLLOWING USING CONTROL RODS AND BORATION:
C a) Rod Bank LO/LOLO Limit a) Do the following to ensure Tech Spec 3.1.1 Annunciators NOT LIT and 3.1.6 compliance: C 1) Immediately initiate boration until the required shutdown margin is restored. C 2) Restore Control Rods to above the Insertion Limits within 30 minutes. C b) A.F.D. Monitor C b) IF alarm is valid, THEN restore AFD within limits Annunciator Panel A H OR reduce thermal power to <50% RTP within NOT LIT 30 minutes in accordance with Tech Spec 3.2.3.
NUMBER PROCEDURE TITLE REVISION 19 1-AP-2.2 FAST LOAD REDUCTION PAGE 4 of 18 H_STEP_[H ACTION/EXPECTEDRESPONSE I RESPONSENOTOBTAINED I CAUTION:
- IF fast load reduction CANNOT be controlled OR if the criteria in ATTACHMENT 3 are exceeded, THEN the ramp SHALL be terminated by performing 1-E-O, REACTOR TRIP OR SAFETY INJECTION OR 1-AP-2.1, TURBINE TRIP WITHOUT REACTOR TRIP REQUIRED, as applicable.
- Auxiliary Steam pressure must be monitored and maintained within specification to both Units in service Air Ejectors.
- Do NOT rely on U0980, Power Calorimetric, to represent real power during a Unit ramp.
U0980 lags real power due to its filtered flow input. Use alternate power indications U1231, Power 1 Mm Avg UFM, U1203, Power 1 Mm Avg FW, and U1201, Power 1 Mm AVG Stm, to monitor real power. Nls, ATs, SG pressure, and First Stage pressure may also be used to monitor real power. NOTE:
- EPIP and notification evaluations may be performed concurrently with the remainder of the procedure.
- Actions not requiring completion of preceding steps may be performed out of sequence at SROs discretion.
- If at any time plant conditions no longer require Fast Load Reduction, then actions should continue with Step 21.
8._ INITIATE NOTIFICATIONS AND EVALUATIONS: a) Notify Energy Supply (MOC) that load reduction has commenced D b) Evaluate EPIP using EPIP-1.O1, EMERGENCY MANAGER CONTROLLING PROCEDURE (STEP 8 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 19 1-AP-2.2 FAST LOAD REDUCTION PAGE 5 of 18 -_STEP ACTION/EXPECTED RESPONSE [ RESPONSE NOT OBTAINED
- 8. INITIATE NOTIFICATIONS AND EVALUATIONS: (Continued)
D c) Evaluate other notifications using VPAP-2802, NOTIFICATIONS AND REPORTS d) Notify Chemistry personnel to perform an Isotopic Analysis for power reductions greater than 15 percent in 1 hour 9._ ATAPPROXIMATELY9O% REACTOR POWER, ENSURE AUXILIARY STEAM IS TRANSFERRED FROM UNIT I SECOND POINT EXTRACTION TO ANOTHER SOURCE USING 1-OP-35.1, AUXILIARY STEAM SYSTEM
*10. VERIFY HP TB GLAND STEAM IF pressure cannot be increased, THEN do the PRESSURE (1-MS-PI-131) - following in the HP Turbine enclosure:
INDICATES 1.5 TO 15 PSIG D
- Throttle open 1-MS-I 98, Gland Steam Supply Header I-MS-PCV-120 Bypass Valve AND
- Control pressure on 1-MS-PI-118, HP Turbine Rear Gland Pressure between 1.5 and 5 PSIG 1l._ AT APPROXIMATELY 85% POWER, STOP THE LOW PRESSURE HEATER DRAIN PUMPS
NUMBER PROCEDURE TITLE REVISION 19 1-AP-2.2 FAST LOAD REDUCTION PAGE 6 of 18 H_STEP_[-H ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED 12._ ATAPPROXIMATELY55% REACTOR POWER, DO THE FOLLOWING: a) Verify two Main Feedwater Pumps- a) Continue load reduction. IN SERVICE WHEN less than 50% reactor power, THEN GO TO Step 14. b) Verify unisolated MFW Recirc D b) Open valve as required. Valve OPEN 13._ AT LESS THAN OR EQUAL TO 50% REACTOR POWER, DO THE FOLLOWING: D a) Verifytwo Main Feedwater Pumps- D a) GO TO Step 14. IN SERVICE b) Stop one Main Feedwater Pump, if desired, as follows: D 1) Place both control switches for standby MEW Pump in PTL D 2) Close Discharge MOV for MFW Pump to be stopped D 3) Place both control switches in PTL for MEW Pump to be stopped C 4) Place Aux Lube Oil Pump in HAND for stopped MFW Pump
NUMBER PROCEDURE TITLE REVISION 19 1-AP-2.2 FAST LOAD REDUCTION PAGE 7 of 18 H_STEP_[-H ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED C 5) Return both control switches for standby MEW Pump to AUTO-AFTER-STOP NOTE: If the fast load reduction is due to loss of 2/3 Main Feedwater Pumps, then the plant may be stabilized at 50% power, if desired, by continuing actions with Step 21. 14._ STOP THE FOLLOWING HP HEATER DRAIN PUMPS: C
- 1-SD-P-IA, A HP HEATER DRAIN PUMP C
- 1-SD-P-1B, B HP HEATER DRAIN PUMP NOTE: Annunciator Panel P-H7, AMSAC OPERATIONAL BYPASS, has a 6-minute time delay associated with it.
15._ ATAPPROXIMATELY35% REACTOR POWER, DO THE FOLLOWING: a) Disable AMSAC as follows: C 1) Verify Panel P H-7, C 1) WHEN annunciator is lit, THEN perform AMSAC OPERATIONAL Step 15a2 and Step 15a3. BYPASS LIT-C Continue with Step 15b. C 2) PIaceAMSACRESET! NORMAL/BYPASS switch in BYPASS yposc J4y14 C 3) Verify Panel A F-7, AMSAC MAN BYP LIT - (STEP 15 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 19 1-AP-2.2 FAST LOAD REDUCTION PAGE 8 of 18 -_STEP f ACTION/EXPECTEDRESPONSE I RESPONSENOTOBTAINED I
- 15. AT APPROXIMATELY 35%
REACTOR POWER, DO THE FOLLOWING: (Continued) D b) Reactor Power will be reduced to D b) GO TO Step 21. less than 30% c) Remove Reheat Steam System as follows: D 1) At Reheater Control Panel, slowly decrease Manual Valve Positioner knob to Zero D 2) Push RESET button and E1 2) Locally close valves. verify MSR FCVs - CLOSED D 3) Stop 1-SD-P-i C, C HP HEATER DRAIN PUMP
- 4) Initiate alignment of MSRs for startup using 1-OP-28.3, STARTUP OF THE MOISTURE SEPARATOR REH EATERS
NUMBER PROCEDURE TITLE REVISION 19 1-AP-2.2 FAST LOAD REDUCTION PAGE 9 of 18 4_STEP_[1 ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED CAUTION: Each RCS Loop Tave MUST be 541 °F in accordance with Tech Spec 3.4.2.
- 16. ATAPPROXIMATELY25% REACTOR POWER, DO THE FOLLOWING:
D a) Verify only one Main Feedwater a) Stop one Main Feedwater Pump as follows: Pump running
- 1) Place both control switches for standby MFW Pump in PTL.
- 2) Ensure Discharge MOV for standby pump is closed.
- 3) Close Discharge MOV for MFW Pump to be stopped.
D 4) Place both control switches in PTL for MFW Pump to be stopped. D 5) Place Aux Lube Oil Pump in HAND for stopped MFW Pump. D 6) Return both control switches for standby MEW Pump to AUTO-AFTER-STOP. D b) Transfer Feedwater Control to Bypass FCVs using ATTACHMENT 2, MAIN FEED CONTROL TRANSFER
NUMBER PROCEDURE TITLE REVISION 19 1-AP-2.2 FAST LOAD REDUCTION PAGE 10 of 18 H_STEP_[ ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED 17._ AT ABOUT 18% REACTOR POWER EJ H Steam Dumps not available, THEN GO TO (170 MWE), TRANSFER STEAM Step 18. FLOW TO STEAM DUMPS: a) Verify Steam Dump Demand a) Do one of the following: indicates zero Stabilize temperature: D a. Reduce ramp rate and control Tavg/Tref until Steam Dump Demand indicates zero. D b. Continue with Step 17b. OR
- Transfer Steam Dumps to Steam Pressure Mode in manual:
- a. Put Steam Dump controller to MANUAL.
D b. Match Steam Dump controller output to demand indicated on Tl-1408. D c. Put Mode Selector switch to STEAM PRESS. D d. Place Steam Dump Controller in AUTO.
- e. Continue with Step 17c.
(STEP 17 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 19 1-AP-2.2 FAST LOAD REDUCTION PAGE 11 of 18 -j_STEP jj ACTION/EXPECTEDRESPONSE I H RESPONSENOTOBTAINED I
- 17. AT ABOUT 18% REACTOR POWER (170 MWE), TRANSFER STEAM FLOW TO STEAM DUMPS:
(Continued) b) Transfer Steam Dump System to Steam Pressure Mode: D 1) Place both Steam Dump Interlock switches in OFF! RESET D 2) Place Steam Dump Controller in MANUAL D 3) Place Condenser Steam Dump Mode Selector switch in STEAM PRESS D 4) Verify or reduce Steam Dump to 0% D 5) Place Steam Dump Controller in AUTO D 6) Verify Steam Dump at 0% D 7) Place both Steam Dump Interlock switches in ON EJ c) Verify Tref at 553 °F D c) WHEN Tref reaches 553 ° F, THEN perform Step 17d. D Continue with Step 18. (STEP 17 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 19 1-AP-2.2 FAST LOAD REDUCTION PAGE 12 of 18 STEP ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED
- 17. AT ABOUT 18% REACTOR POWER (170 MWE), TRANSFER STEAM FLOW TO STEAM DUMPS:
(Continued) d) Check Steam Dump response: D 1) Verify no Control Rod motion (O D 2) Place Control Rod Mode I1a Selector switch in MANUAL -i D 3) Maintain Tavg greater than Tref during Turbine rampdown to allow Steam 47 Header Pressure to slowly increase to 1005 psig D 4) Verify Steam Header D 4) WHEN pressure is approximately 1005 psig, pressure AT 1005 PSIG
- THEN verify Steam Dumps respond to control Steam Header pressure.
D Continue with Step 18.
NUMBER PROCEDURE TITLE REVISION 19 1-AP2.2 FAST LOAD REDUCTION PAGE 13 of 18 -_STEP_[1 ACTION/EXPECTEDRESPONSE RESPONSENOTOBTAINED I NOTE: IF both IR NIs have failed low, THEN before power is reduced below P-b (10 percent power), actions must be taken using 1-AP-4.2 to prevent a Source Range High Flux Reactor Trip.
- 18. ATAPPROXIMATELY 15% REACTOR POWER (140 MWE), DO THE FOLLOWING:
1J a) Verify Panel P G-1, C-5 INTLK a) WHEN annunciator is lit, THEN continue with TURB IMP PRESS <15% AUTO this procedure. ROD BLK LIT-C b) Verify no Control Rod motion C c) Verify Control Rod Mode Selector C c) Place Control Rod Mode Selector switch in switch in MANUAL MANUAL. C d) Place Turbine Control in HOLD AND match Reference I Setter values C e) Place Turbine Load Control in . IMP-OUT by depressing the IMP-OUT pushbutton C f) Continue to lower Main Turbine Load r C g) Verify all Turbine Drain Valves automatically open
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 65. 02.2.37 065/NEW/1H14/3.6/4.6//
Given the following conditions:
- Unit 1 was initially at 100% power.
- PRZR Pressure Control Transmitter, 1-RC-PT-1445, failed HIGH.
- The crew placed the control switch for 1-RC-PCV-1456, PRZR PORV, in CLOSE.
- RCS pressure subsequently returned to 2235 psig.
- PRT pressure and temperature have stabilized.
Which ONE of the following identifies the Technical Specification implications of these plant conditions? 1-RC-PCV-1456 and 1-RC-MOV-1 535, PRZR PORV Block Valve, A. is operable; MUST be closed to comply with Technical Specifications. B. is operable; is NOT required to be closed to comply with Technical Specifications. C. is NOT operable; MUST be closed to comply with Technical Specifications. D. is NOT operable; is NOT required to be closed to comply with Technical Specifications.
- a. Incorrect. First part is true TS only requires the valve be capable of manually cycling which it is.
Second part is incorrect but plausible, since under certain conditions (such as excessive leakage) the block valve is required to be closed. The candidate who does not have detailed understanding of the Tech Specs which discusses reasons requiring isolation, vice permitting closure of the block valve as a matter of prudent judgement may default to this distractor.
- b. Correct. First part is correct as noted above. Second part is also correct; because this is only a control system related malfunction TS allow continued operation as long as the PORV is under manual control. While there is nothing that prohibits closure of the block valve, there is no requirement to do so.
- c. Incorrect. First part is incorrect but plausible since as previously discussed the candidate who does not have detailed knowledge of the Tech Specs might conclude that since the automatic function is no longer working that the PORV would thus be inoperable. Second part is incorrect as discussed in Distractor a.
- d. Incorrect. First part is incorrect as discussed in Distractor c. Second part is correct as discussed in distractor b.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Equipment Control Ability to determine operability and/or availability of safety related equipment. (CFR: 41.7 / 43.5 / 45.12) Tier: 3 Technical
Reference:
TS 3.4.11 & bases Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
NUCLEAR DESGN NFORMATON PORTAL
- Pressurizer PORVs 3.4.11 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.11 Pressurizer Power Operated Relief Valves (PORVs)
LCO 3.4.11 Each PORV and associated block valve shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTIONS
- NOTE -
Separate Condition entry is allowed for each PORV and each block valve. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more PORVs A.1 Restore backup 14 days inoperable due to nitrogen supply to inoperable backup OPERABLE status. nitrogen supply and capabl e of being manually cycled. B. One or more PORVs B.1 Close and maintain 1 hour inoperable for reason power to associated other than Condition A block valve. and capable of being manually cycled. C. One PORV inoperable C.1 Close associated block 1 hour and not capable of valve. being manually cycled. AND C.2 Remove power from 1 hour associated block val ye. AND C.3 Restore PORV to 72 hours OPERABLE status. North Anna Units 1 and 2 3 .4. 11-1 Amendments 231/212
NUCLEAR DESGN NFORMATON PORTAL Pressurizer PORVs 3.4.11 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. One block valve NOTE inoperable. Required Action D.1 and D.2 do not apply when block valve
\ is inoperable solely as a result of complying with I Required Action C.2.
A A y C, *i ,7 *JZ D.1 Place associated PORV 1 hour k in manual control I U 7 AND D.2 Restore block valve to 72 hours OPERABLE status. E. Required Action and E.1 Be in MODE 3. 6 hours associated Completion Time of Condition A, AND B, C, or D not met. E.2 Be in MODE 4. 12 hours F. Two PORVs inoperable F.1 Be in MODE 3. 6 hours and not capable of being manually cycled. AND F.2 Be in MODE 4. 12 hours North Anna Units 1 and 2 3 .4. 11-2 Amendments 231/212
NUCLEAR DESIGN INFORMAIION PORTAL Pressurizer PORVs 3.4.11 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME G. Two block valves G.1 NOTE inoperable. Required Action G.1 does not apply when block valve is inoperable solely as a result of complying with Required Action C.2. Restore one block 2 hours valve to OPERABLE status. H. Required Action and H.1 Be in MODE 3. 6 hours associated Completion Time of Condition G AND not met. H.2 Be in MODE 4. 12 hours SURVEILLANCE_REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1 Verify PORV backup nitrogen supply pressure 7 days is within limit. SR 3.4.11.2 NOTES
- 1. Not required to be performed with block valve closed in accordance with the Required Actions of this LCO.
- 2. Only required to be performed in MODES 1 and 2.
Perform a complete cycle of each block 92 days valve. North Anna Units 1 and 2 3.4.11-3 Amendments 231/212
Pressurizer PORVs 3.4.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.3 NOTE Only required to be performed in MODES 1 and 2. Perform a complete cycle of each PORV. 18 months SR 3.4.11.4 Perform a complete cycle of each solenoid 18 months control valve and check valve on the accumulators in PORV control systems. North Anna Units 1 and 2 3.4.11-4 Amendments 231/212
NUCLEAR DESIGN NFORMA11ON PORTAL Pressurizer PORVs B 3.4.11 B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.11 Pressurizer Power Operated Relief Valves (PORVs) BASES BACKGROUND The pressurizer is equipped with two types of devices for pressure relief: pressurizer safety valves and PORVs. The PORVs are air or nitrogen operated valves that are controlled to open at a set pressure when the pressurizer pressure increases and close when the pressurizer pressure decreases. The PORVs may also be manually operated from the control room. Block valves, which are normally open, are located between the pressurizer and the PORVs. The block valves are used to isolate the PORVs in case of excessive leakage or a stuck open PORV. Block valve closure is accomplished manually using controls in the control room. A stuck open PORV is, in effect, a small break loss of coolant accident (LOCA). As such, block valve closure terminates the RCS depressurization and coolant inventory loss. The PORVs and their associated block valves may be used by unit operators to depressurize the RCS to recover from certain transients if normal pressurizer spray is not available. Additionally, the series arrangement of the PORVs and their block valves permit performance of surveillances 7 on the valves during power operation. The PORVs may also be used for feed and bleed core cooling in the case of multiple equipment failure events that are not within the design basis, such as a total loss of feedwater. The PORVs, their block valves, and their controls are powered from the emergency buses that normally receive power from offsite power sources, but are also capable of being powered from emergency power sources in the event of a loss of offsite power. The PORVs are air operated valves and normally are provided motive force by the Instrument Air System. A backup, nitrogen supply for the PORVs is also available. Two PORVs and their associated block valves are powered from two separate safety trains (Ref. 1). The unit has two PORVs, each having a relief capacity of 210,000 lb/hr at 2335 psig. The functional design of the PORVs is based on maintaining pressure below the Pressurizer (conti nued) North Anna Units 1 and 2 B 3.4.11-1 Revision 0
NUCLEAR DESIGN INFORMATION PORTAL Pressurizer PORVs B 3.4.11 BASES BACKGROUND PressureHigh reactor trip setpoint following a step (continued) reduction of 50% of full load with steam dump. In addition, the PORVs minimize challenges to the pressurizer safety valves and also may be used for low temperature overpressure protection (LTOP). See [CO 3.4.12, Low Temperature Overpressure Protection ([TOP) System. APPLICABLE Unit operators employ the PORVs to depressurize the RCS in SAFETY ANALYSES response to certain unit transients if normal pressurizer spray is not available. For the Steam Generator Tube Rupture (SGTR) event, the safety analysis assumes that manual operator actions are required to mitigate the event. A loss of offsite power is assumed to accompany the event, and thus, normal pressurizer spray is unavailable to reduce RCS pressure. The PORVs are assumed to be used for RCS depressurization, which is one of the steps performed to equalize the primary and secondary pressures in order to terminate the primary to secondary break flow and the radioactive releases from the affected steam generator. The PORVs are also modeled in safety analyses for events that result in increasing RCS pressure for which departure from nucl eate boiling ratio (DNBR) cri ten a are cri ti cal (Ref. 2). By assuming PORV actuation, the primary pressure remains below the high pressurizer pressure trip setpoint; thus, the DNBR calculation is more conservative. As such, this actuation is not required to mitigate these events, and PORV automatic operation is, therefore, not an assumed safety function. Pressurizer PORVs satisfy Criterion 3 of 10 CFR 50.36(c) (2) (ii). [CO The [CO requires the PORVs and their associated block valves to be OPERABLE for manual operation to mitigate the effects associated with an SGTR. By maintaining two PORVs and their associated block valves OPERABLE, the single failure criterion is satisfied. An OPERABLE block valve may be either open and energized with the capability to be closed, or closed and energized with the capability to be opened, since the required safety function is accomplished by manual operation. Although typically open to allow PORV operation, the block valves may be OPERABLE when closed to isolate the flow path of an inoperable PORV (conti nued) North Anna Units 1 and 2 B 3.4.11-2 Revision 0
NUCLEAR DESIGN NFORMATON PORTAL Pressurizer PORVs B 3.4.11 BASES L CO that is capable of being manually cycled (e.g., as in the (conti nued) case of excessive PORV leakage). Similarly, isolation of an OPERABLE PORV does not render that PORV or block valve inoperable provided the relief function remains available with manual action. I An OPERABLE PORV is required to be capable of manually pV opening and closing, and not experiencing excessive seat leakage. Excessive seat leakage, although not associated with a specific acceptance criteria, exists when conditions dictate closure of the block valve to limit leakage to within (\9] LCO 3.4.13, RCS Operational Leakage. 11 Satisfying the [CO helps minimize challenges to fission product barriers. APPLICABILITY In MODES 1, 2, and 3, the PORVs and their associated block valves are required to be OPERABLE to limit the potential for a small break LOCA through the flow path and for manual operation to mitigate the effects associated with an SGTR. The PORVs are also required to be OPERABLE in MODES 1, 2, and 3 for manual actuation to mitigate an SGTR event. Imbalances in the energy output of the core and heat removal by the secondary system can cause the RCS pressure to increase to the PORV opening setpoint. The most rapid
\A increases will occur at the higher operating power and pressure conditions of MODES 1 and 2.
C) Pressure increases are less prominent in MODE 3 because the core input energy is reduced, but the RCS pressure is high. Therefore, the LCO is applicable in MODES 1, 2, and 3. The LCO is not applicable in MODES 4, 5, and 6 with the reactor vessel head in place when both pressure and core energy are decreased and the pressure surges become much less significant. LCO 3.4.12 addresses the PORV requirements in these MODES. ACTIONS Note 1 has been added to clarify that all pressurizer PORVs are treated as separate entities, each with separate Completion Times (i.e., the Completion lime is on a component basis). North Anna Units 1 and 2 B 3.4.11-3 Revision 0
NUCLEAR DESIGN NFORMATON PORTAL
- Pressurizer PORVs B 3.4.11 BASES ACTIONS A.1 (conti nued)
The PORVs are provided normal motive force by the Instrument Air system and have a backup nitrogen supply. If the backup nitrogen supply is inoperable, the PORV5 are still capable of being manually cycled provided the Instrument Air system is available. The Instrument Air system is highly reliable and the likelihood of its being unavailable during a demand for PORV actuation is low enough to justify a 14 day Completion Time for return of the backup nitrogen supply to OPERABLE status. B.1 PORVs may be inoperable and capable of being manually cycled (e.g., excessive seat leakage). In this Condition, either the PORVs must be restored or the flow path isolated within 1 hour. The associated block valve is required to be closed, but power must be maintained to the associated block valve, since removal of power would render the block valve inoperable. This permits operation of the unit until the next refueling outage (MODE 6) so that maintenance can be performed on the PORVs to eliminate the problem condition. Quick access to the PORV for pressure control can be made when power remains on the closed block valve. The Completion Time of 1 hour is based on unit operating experience that has shown that minor problems can be corrected or closure accomplished in this time period. C.1, C.2, and C.3 If one PORV is inoperable and not capable of being manually cycled, it must be either restored, or isolated by closing the associated block valve and removing the power to the associated block valve. The Completion Time of 1 hour is reasonable, based on challenges to the PORVs during this time period, and provides the operator adequate time to correct the situation. If the inoperable valve cannot be restored to being capable of being manually cycled (permitting entry into Condition B), or OPERABLE status, it must be isolated within the specified time. Because there is one PORV that remains OPERABLE, an additional 72 hours is provided to restore the inoperable PORV to OPERABLE status. If the PORV cannot be restored within this additional time, the unit must be brought to a MODE in which the LCO does not apply, as required by Condition E. North Anna Units 1 and 2 B 3.4.11-4 Revision 0
NUCLEAR DESIGN NFORMAflON PORTAL Pressurizer PORVs B 3.4.11 BASES ACTIONS D.1 and D.2 (conti nued) If one block valve is inoperable, then it is necessary to either restore the block valve to OPERABLE status within the Completion Time of 1 hour or place the associated PORV in manual control. The prime importance for the capability to close the block valve is to isolate a stuck open PORV. Therefore, if the block valve cannot be restored to OPERABLE status within 1 hour, the Required Action is to place the PORV in manual control to preclude its automatic opening for an overpressure event and to avoid the potential for a stuck open PORV at a time that the block valve is inoperable. The Completion Time of 1 hour is reasonable, based on the small potential for challenges to the system during this time period, and provides the operator time to correct the situation. Because at least one PORV remains OPERABLE, the operator is permitted a Completion Time of 72 hours to restore the inoperable block valve to OPERABLE status. The time allowed to restore the block valve is based upon the Completion Time for restoring an inoperable PORV in Condition C, since the PORVs may not be capable of mitigating an event if the inoperable block valve is not full open. If the block valve is restored within the Completion Time of 72 hours, the PORV may be restored to automatic operation. If it cannot be restored within this additional time, the unit must be brought to a MODE in which the [CO does not apply, as required by Condition E. The Required Actions D.1 and D.2 are modified by a Note stating that the Required Actions do not apply if the sole reason for the block valve being declared inoperable is as a result of power being removed to comply with another Required Action. In this event, the Required Actions for inoperable PORV(s) (which require the block valve power to be removed once it is closed) are adequate to address the condition. While it may be desirable to also place the PORV(s) in manual control, this may not be possible for all causes of Condition C entry with PORV(s) inoperable and not capable of being manually cycled (e.g., as a result of failed control power fuse(s) or control switch malfunction(s).) E.1 and E.2 If the Required Action of Condition A, B, C, or D is not met, then the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 4 within (conti nued) North Anna Units 1 and 2 B 3.4.11-5 Revision 0
NUCLEAR DESGN INFORMATION PORTAL Pressurizer PORVs B 3.4.11 BASES ACTIONS E.1 and E.2 (continued) 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. In MODE 4, automatic PORV OPERABILITY is required. See LCO 3.4.12. F.1 and F.2 If more than one PORV is inoperable and not capable of being manually cycled, then the unit must be brought to a MODE in which the [CO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. In MODE 4, automatic PORV OPERABILITY is required. See [CO 3.4.12. G.1 If two block valves are inoperable, it is necessary to restore at least one block valve within 2 hours. The Completion Time is reasonable, based on the small potential for challenges to the system during this time and provide the operator time to correct the situation. The Required Action G.1 is modified by a Note stating that the Required Action does not apply if the sole reason for the block valve being declared inoperable is as a result of power being removed to comply with another Required Action. In this event, the Required Action for inoperable PORV (which requires the block valve power to be removed once it is closed) is adequate to address the condition. While it may be desirable to also place the PORV in manual control, this may not be possible for all causes of Condition C entry with PORV inoperable and not capable of being manually cycled (e.g., as a result of failed control power fuse(s) or control switch malfunction(s)). H.1 and H.2 If the Required Actions of Condition G are not met, then the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at (conti nued) North Anna Units 1 and 2 B 3.4.11-6 Revision 0
NUCLEAR DESIGN NFORMATION PORTAL Pressurizer PORVs B 3.4.11 BASES ACTIONS H.l and H.2 (continued) least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. In MODE 4, automatic PORV OPERABILITY is required. See [CO 3.4.12. SURVEILLANCE SR 3.4.11.1 REQUIREMENTS SR 3.4.11.1 requires verification that the pressure in the PORV backup nitrogen system is sufficient to provide motive force for the PORVs to cope with a steam generator tube rupture coincident with loss of the containment Instrument Air system. The Frequency of 7 days is based on operating experience. SR 3.4.11.2 Block valve cycling verifies that the valve(s) can be opened and closed if needed. The basis for the Frequency of 92 days is the ASME Code (Ref. 3). This SR is modified by two Notes. Note 1 modifies this SR by stating that it is not required to be performed with the block valve closed, in accordance with the Required Actions of this LCO. Opening the block valve in this condition increases the risk of an unisolable leak from the RCS since the PORV is already inoperable. Note 2 modifies this SR to allow entry into and operation in MODE 3 prior to performing the SR. This allows the test to be performed in MODE 3 under operating temperature and pressure conditions, prior to entering MODE 1 or 2. SR 3.4.11.3 SR 3.4.11.3 requires a complete cycle of each PORV. Operating a PORV through one complete cycle ensures that the PORV can be manually actuated for mitigation of an SGTR. This testing is performed in MODES 3 or 4 to prevent possible RCS pressure transients with the reactor critical The Frequency of 18 months is based on a typical refueling cycle and industry accepted practice. (conti nued) North Anna Units 1 and 2 B 3.4.11-7 Revision 0
NUCLEAR DESIGN INFORMATION PORTAL Pressurizer PORVs B 3.4.11 BASES SURVEILLANCE SR 3.4.11.3 (continued) REQUIREMENTS The Note modifies this SR to allow entry into and operation in MODE 3 prior to performing the SR. This allows the test to be performed in MODE 3 under operating temperature and pressure conditions, prior to entering MODE 1 or 2. SR 3.4.11.4 Operating the solenoid control valves and check valves on the accumulators ensures the PORV control system actuates properly when called upon. The Frequency of 18 months is based on a typical refueling cycle and the Frequency of the other Surveillances used to demonstrate PORV OPERABILITY. REFERENCES 1. Regulatory Guide 1.32, February 1977.
- 2. UFSAR, Section 15.4.
- 3. ASME Code for Operation and Maintenance of Nuclear Power Plants.
North Anna Units 1 and 2 B 3.4.11-8 Revision 0
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 66. G2.3.11 066/NEWI/H13/3 .8/4.3/!
Given the following conditions:
- Unit 1 is at 50% power and ramping up at 0.3%/minute.
- 1-OW-P-iD, D Circulating Water pump, trips.
- The crew enters 1-AP-13, Loss of One or More Circulating Water Pumps.
Which ONE of the following identifies the guidance provided by 1-AP-13 for controlling Liquid Waste releases? A. Liquid Waste releases must be immediately secured, and Chemistry Department approval must be obtained prior to resuming them. B. Liquid Waste releases must be immediately secured, and Health Physics Department approval must be obtained prior to resuming them. C. Inform Chemistry Department of Circulating Water Pumps status, and they will determine if Liquid Waste releases are required to be secured. D. Inform Health Physics Department of Circulating Water Pumps status, and they will determine if Liquid Waste releases are required to be secured.
- a. Incorrect. First part incorrect but plausible since some procedures like 0-AP-lO do have you conservatively stop releases. Second part also incorrect but plausible since the subject department is associated with sampling and analysis.
- b. Incorrect. First part incorrect but plausible as discussed above. Second part is correct this department has approval authority for release permits.
- c. Incorrect. Although there is no immediate requirement to stop releases as mentioned above Chemistry although plausible is not the approval authority.
- d. Correct. Health Physics Department will be contacted since they are the approval authority and based on CW pump status and release permit requirements they will advise operations as to whether releases may continue, or are required to be secured.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Radiation Control Ability to control radiation releases. (CFR: 41.11 /43.4/45.10) Tier: 3 Technical
Reference:
1-AP-13 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
e o
Dotiiion NORTH ANNA POWER STATION ABNORMAL PROCEDURE NUMBER PROCEDURE TITLE REVISION 15 1 -AP-1 3 LOSS OF ONE OR MORE CIRCULATING WATER PUMPS (WITH ONE ATTACHMENT) PAGE 1 of 8 PURPOSE To provide instructions to follow in the event that one or more Circulating Water Pumps are lost. ENTRY CONDITIONS This procedure is entered when any of the following conditions exist:
- TRIP indication on breaker control switches on the Circulating Water Control Panel, or
- One or more Circulating Water Pump Motors read zero amps, or
- One or more Circulating Water Pump Motors degraded as indicated by low amps, or
- Annunciator Panel B A-5, CW PP lA-lB-iC-iD AUTO TRIP, is LIT CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 15 1 -AR-i 3 LOSS OF ONE OR MORE CIRCULATING WATER PUMPS PAGE 2 of 8 STEP ACTION/EXPECTEDRESPONSE I I RESPONSENOTOBTAINED I 1 ] VERIFY AT LEAST TWO D GO TO 1-E-O, REACTOR TRIP OR WATER CIRCULATING PUMPS - SAFETY INJECTION, while continuing with this RUNNING procedure.
*2 VERIFY CONDENSER VACUUM: D Initiate 1-AP-14, LOW CONDENSER VACUUM, while continuing with this procedure.
- 3.5 INCHES HG ABS OR LESS AND D
- STABLE
- 3. CHECK CIRCULATING WATER Manually or locally close affected MOV:
PUMP DISCHARGE MOV ON ANY NON-RUNNING PUMP CLOSED - D
- 1-CW-MOV-100A
- 1-CW-MOV-100B D
- 1-CW-MOV-100C D
- 1-CW-MOV-100D 4._ CHECK ANY CIRCULATING WATER D GO TO Step 5.
PUMP DEGRADED AS INDICATED I BYLOWAMPS D a) Place affected Circulating Water Pump control switch in PTL D b) Verify affected pump discharge b) Manually or locally close affected MOV: MOV closed
/ C
- 1-CW-MOV-100A C
- 1-CW-MOV-100B C
- 1-CW-MOV-100C C
- 1-CW-MOV-100D
NUMBER PROCEDURE TITLE REVISION 15 1 -AP-1 3 LOSS OF ONE OR MORE CIRCULATING WATER PUMPS PAGE 3 of 8 -j_STEP ACTION/EXPECTEDRESPONSE I I RESPONSENOTOBTAINED I
- 5. CHECK OPERATING CIRCULATING Reduce operating Circulating Water Pump amps WATER PUMP AMPS LESS THAN by doing one of the following, while continuing with
/ 340 AMPS this procedure:
- Throttle discharge MOVs using 1-OP-48.2, Operation of Circulating Water System.
OR
- Reduce number of water boxes in service using the following applicable procedure:
D
- A Water Box 1 -MOP-48.30
- BWaterBox-1-MOP-48.31 D
- C Water Box 1-MOP-48.32 D
- D Water Box MOP-48.33 VERIFY BEARING COOLING IF Bearing Cooling is aligned Lake-to-Lake, THEN SYSTEM SUPPORT: do the following:
D
- Circ Water Intake Tunnel Full -
- Place 1 -WT-P-25 in service using 1 -OP-48.4, LIGHT LIT OPERATION OF AUXILIARY FLASH EVAPORATOR PUMP 1-WT-P-25.
C
- At least one Circulating Water pump-RUNNING C
- Align Bearing Cooling to Tower-to-Tower using 1-OP-50.2, OPERATION OF THE BEARING COOLING WATER SYSTEM.
C GO TO 1-AP-19, LOSS OF BEARING COOLING WATER, while continuing with this procedure.
NUMBER PROCEDURE TITLE REVISION 15 1 -AP-1 3 LOSS OF ONE OR MORE CIRCULATING WATER PUMPS PAGE 4 of 8 ACTION! EXPECTED RESPONSE I I RESPONSE NOT OBTAINED
- 7. DETERMINE IF SECURING LIQUID ,
WASTE RELEASES REQUIRED: iJ_ I D a) Have HP Count Room determine if securing liquid waste releases is cZ required based on Circulating ( ) ** Water system status D b) Health Physics has determined , D b) GO TO Step 9. securing liquid waste releases is required 8._ SECURE LIQUID WASTE RELEASES: a) Verify SG Blowdown Trip Valves - a) Do the following, as required: CLOSED: D
- Manually close SG Blowdown Trip valves.
D
- Unit 1 SG Blowdown OR D
- Unit 2 SG Blowdown
- Locally close the following:
D
- SG Blowdown HCVs for Low Capacity SG Blowdown
- SG Blowdown FCVs for High Capacity SG Blowdown b) Send an operator to locally close the following:
D
- 1-BD-1005, SG BD flash tk to SG BD Flash Tk Drn CIrs Isol Vv D
- 2-BD-182, SG BD Flash Tk To SG BD Flash Tk Drn CIrs Isol Vv (STEP 8 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 15 1 -AP-1 3 LOSS OF ONE OR MORE CIRCULATING WATER PUMPS PAGE 5 of 8 H_STEP_j ACTION/EXPECTEDRESPONSE I RESPONSENOTOBTAINED
- 8. SECURE LIQUID WASTE RELEASES: (Continued) c) Stop the following pumps to prevent clarifier system overflow:
- Contaminated Drain Tank Pumps:
C
- 1-LW-P-6A C
- 1-LW-P-6B
- Low Level Liquid Waste Tank Pumps:
C
- 1-LW-P-lA C
- 1-LW-P-lB
- Boron Recovery Test Tank Pumps:
C
- 1-BR-P-5A C
- 1-BR-P-SB d) Send an operator to stop Containment Matt Sump Pumps:
- 1) Locally open SOV breakers:
C
- Unit 1 1-EP-DB-2, Breakers 6 AND 7 C
- Unit 2 2-EP-DB-5, Breakers 6 AND 7 C 2) Verify Pumps are stopped C 2) Manually isolate Service Air to Pump(s).
(STEP 8 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 15 1 -AP-1 3 LOSS OF ONE OR MORE CIRCULATING WATER PUMPS PAGE 6 of 8 H_STEP [H ACTION/EXPECTEDRESPONSE I -J RESPONSENOTOBTAINED I
- 8. SECURE LIQUID WASTE RELEASES: (Continued)
D e) Verify Liquid Waste release D e) Place 1-LW-PCV-115 control in HAND and secured by verifying close. 1-LW-PCV-115 IN HAND CONTROL AND CLOSED
- 9. VERIFY LPTURBINE RUPTURE Secure condenser air ejectors using 1-OP-36.2, DISCS INTACT
- MAIN CONDENSER AIR EJECTOR SYSTEM:
D
- WHEN condenser vacuum reaches zero, THEN secure Gland Steam System using 1-OP-39.1, Gland Seal Steam System.
D
- Enter Condition Report to replace rupture discs.
1O. CHECK HOTWELL TEMPERATURE -
- Bypass the Powdex system using 1-OP-30.2, LESS THAN 130 °F Powdex System.
D
- IF Condensate Recirc is required, THEN adjust 1 -CN-FCV-1 07 setpoint as necessary OR fail open using 1-OP-30.1, Operation of Condensate System.
NUMBER PROCEDURE TITLE REVISION 15 1 -AP-1 3 LOSS OF ONE OR MORE CIRCULATING WATER PUMPS PAGE 7 of 8 ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED 11._ VERIFY LIQUID WASTE RELEASE Do the following: PATHS AVAILABLE: 0
- Verify liquid waste release forms 0 a) Obtain Health Physics concurrence on alternate in effect Liquid Waste Release paths.
D
- Re-establish liquid waste Gb) Initiate liquid waste release to Unit 2 discharge releases using O-OP-22.11, 4 tunnel using O-OP-22.11, Releasing Radioactive Releasing Radkactive Liquid Liquid Waste.
Wastecjequired
- Re-establish Unit 1 Blowdown, as required:
0
- High Capacity SG Blowdown using 1-OP-32.3, High Lkx Capacity Steam Generator Blowdown System Operation p .i .
OR (.. U* 0
- Low Capacity SG Blowdown using 1-OP-32.1, Low Capacity Steam Generator i2 . .
Blowdown System C. C-Re-establish Unit 2 Blowdown, .L. .. - as required:
,JI
() 0
- High Capacity SG Blowdown using 2-OP-32.3, High /
Capacity Steam Generator Blowdown System Operation I L- I OR 0
- Low Capacity SG Blowdown using 2-OP-32.1, Utilizing Steam Generator Blowdown Tank 2-BD-TK-1
NUMBER PROCEDURE TITLE REVISION 15 1 -AR-i 3 LOSS OF ONE OR MORE CIRCULATING WATER PUMPS PAGE 8 of 8 STEP_[j ACTION/EXPECTEDRESPONSE RESPONSENOTOBTAINED I CAUTION: If the loss of Circulating Water Pumps was caused by SI/CDA load shed, then O-AP-47, UNIT OPERATION DURING OPPOSITE UNIT EMERGENCY, must be used for instructions on resetting Sl/CDA load shed. Circulating Water Pump Breaker control switches must be placed to TRIP position before load shed is reset to prevent a possible auto start of Circulating Water Pumps.
- 12. DETERMINE STATUS OF CIRCULATING WATER PUMP(S):
D a) Condition Report(s) - ENTERED D a) Enter Condition Report(s). D b) Affected Circulating Water Pump(s) D b) Continue with other procedures in effect.
- AVAILABLE D WHEN affected pumps are available, THEN GOTO Step 13.
13._ RESTORE CIRCULATING WATER SYSTEM USING 1-OP-48.2, OPERATION OF THE CIRCULATING WATER SYSTEM
- 14. RETURN TO PROCEDURE AND STEP IN EFFECT
- END -
NUMBER ATTACHMENT TITLE ATTACHMENT 1-AP-13 1 REFERENCES REVISION PAGE 15 lofi
- 11715-FM-77, CIRCULATING WATER SYSTEM
- DCP 88-03-1, TURBINE TRIP - REACTOR TRIP, Setpoint change to P-8
- DCP 94-003, Steam Generator Blowdown System Upgrades Unit 1
- 1 -AP-2.1, TURBINE TRIP WITHOUT REACTOR TRIP REQUIRED
- 1 -OP-32.1, LOW CAPACITY STEAM GENERATOR BLOWDOWN SYSTEM
- 2-OP-32.1, STEAM GENERATOR BLOWDOWN UTILIZING STEAM GENERATOR BLOWDOWN TANK (2-BD-TK-1)
- 1-OP-3a3, HIGH CAPACITY STEAM GENERATOR BLOWDOWN SYSTEM OPERATION
- 2-OP-32.3, HIGH CAPACITY STEAM GENERATOR BLOWDOWN SYSTEM OPERATION
- 1 -OP-36.2, MAIN CONDENSER AIR EJECTORS
- 1-OP-39.1, GLAND SEAL STEAM SYSTEM
- 1 -OP-50.2, OPERATION OF THE BEARING COOLING WATER SYSTEM TOWER-TO-TOWER
- 1-OP-302, POWDEX SYSTEM
- 1 -OP-22.1 1, RELEASING RADIOACTIVE LIQUID WASTE
- 1-OP-48.2, OPERATION OF THE CIRCULATING WATER SYSTEM
- 1-E-0, REACTOR TRIP OR SAFETY INJECTION
- 0-AP-47, UNIT OPERATION DURING OPPOSITE UNIT EMERGENCY
- 1-AP-14, LOW CONDENSER VACUUM
- 1-AP-19, LOSS OF BEARING COOLING WATER
- OP-08-0352, Rewrite AP to Provide Enhancements (Rev 14)
- SOER 07-2, Intake Cooling Water Blockage for CA076323, CR100270 (Rev 14)
NUMBER PROCEDURE TITLE REVISION 62 O-AP-1 LOSS OF ELECTRICAL POWER PAGE 3 of 32 ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED NOTE: If the Low Capacity SG Blowdown system is NOT in service, then it is acceptable to leave High Capacity SG Blowdown in service because loss of power that does affect the High Capacity system will automatically isolate blowdown due to a loss of control power. (Reference 100 and 101)
- 3. CHECK RADIOACTIVE RELEASES FROM BOTH UNITS SECURED:
11 a) Verify Unit 1 Low Capacity SG Blowdown - a) Verify ALL Unit 1 SG Blowdown is ISOLATED isolated: D 1) Verify Unit 1 SG Blowdown Trip Valves CLOSED D 2) IF trip valves are NOT closed, THEN manually close.
- 3) IF trip valves will not close, THEN manually close the following:
D
- SG Blowdown HCVs for Low Capacity SG Blowdown
\
D
- SG Blowdown FCVs for High I,/..( I Capacity SG Blowdown D 4) Send an operator to locally close 1 -BD-1 005, SG BD FLASH TK TO SG BD FLASH TK DRN CLRS ISOL VV (STEP 3 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 62 0-A P-i 0 LOSS OF ELECTRICAL POWER PAGE 4 of 32 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED H_STEP7-_j
- 3. CHECK RADIOACTIVE RELEASES FROM BOTH UNITS SECURED: (Continued)
D b) Verify Unit 2 Low Capacity SG Blowdown - b) Verify ALL Unit 2 SG Blowdown is ISOLATED isolated: D 1) Verify Unit 2 SG Blowdown Trip Valves CLOSED
- 2) IF trip valves are NOT closed, THEN manually close.
- 3) IF trip valves will not close, THEN manually close the following:
C
- SG Blowdown HCVs for Low Capacity SG Blowdown C
- SG Blowdown FCVs for High Capacity SG Blowdown C 4) Send an operator to locally close 2-BD-1 82, SG BD FLASH TK TO SG BD FLASH TK DRN CLRS ISOL VV C c) Verify Liquid Waste release secured by C c) Place 1-LW-PCV-115 control in HAND verifying 1-LW-PCV-115 IN HAND
- and close.
CONTROL AND CLOSED C d) Verify Gaseous Waste release secured by C d) Close valve. verifying 1-GW-FCV-101 CLOSED
- 4. CHECK UNIT 2 EMERGENCY BUSSES - C H Unit 2 is in Modes 1 through 4, THEN GO EITHER ENERGIZED: TO 2-ECA-0.0, LOSS OF ALL AC POWER, while continuing with this procedure C
- 2H 4160-Volt Bus C IF Unit 2 is in Mode 5 or 6, THEN GO TO Step 5.
C
- 2J 4160-Volt Bus
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 67. G2.3.13 067/BANKJNAPS/H/3/3 .4/3.8!!
Which ONE of the following identifies the posting requirements, AND includes the MINIMUM entry requirements for an area accessible to individuals, in which radiation levels could result in an individual receiving a deep dose equivalent in excess of 100 mrem in one hour at 30 centimeters from the source? A. High Radiation Area; alarming dosimetry (DAD) OR continuous HP coverage B. High Radiation Area; alarming dosimetry (DAD) AND continuous HP coverage C. Locked High Radiation Area ; alarming dosimetry (DAD) OR continuous HP coverage D. Locked High Radiation Area; alarming dosimetry (DAD) AND continuous HP coverage
- a. Correct. Posting requirement and access control as given in VPAP-2101. It is acceptable to utilize all three (DAD, HP coverage, and the individual himself have a survey meter) however, the question clearly solicits the minimum requirements, as such or (any one of the three options) demonstrates the candidate correctly understands the lowest allowable level of controls.
- b. Incorrect. First part is correct. Second part is incorrect but plausible since it would seem reasonable (conservative) to a person without detailed knowledge of limits and/or requirements.
- c. Incorrect. Plausible since it would seem reasonable to a person without detailed knowledge of limits and/or requirements.
- d. Incorrect. Plausible since it would seem reasonable to a person without detailed knowledge of limits and/or requirements.
Radiation Control Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. (CFR: 41.12 / 43.4/45.9 / 45.10) Tier: 3 Technical
Reference:
VPAP-2101 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
Station J$ Dominion Administrative Procedure
Title:
Radiation Protection Program Process I Program Owner:Manager Radiological Protection and Chemistry Procedure Number Revision Number Effective Date VPAP-2101 34 On File Revision Summary Revised in response to CA123928, Revise VPAP-2101 to make Steps 6.6.l.e.2 and 6.6.9.a.1 consistent: Old 6.6.9.a. 1: Items that are being free released from the RCA shall be monitored for radioactive contamination by RP. No item may be released for unrestricted use if monitoring indicates the presence of contamination. New 6.6.9.a. 1: Items (other than personal hand-carried items) that are being free released from the RCA shall be monitored for radioactive contamination by RP. No item may be released for unrestricted use if monitoring indicates the presence of contamination. Approvals on File
DOMINION VPAP-2101 REVISION 34 PAGE 43 OF 110 6.3.9 Identification and Control of Radiation Sources
- a. RP shall perform appropriate radiation surveys.
- b. Radiation levels shall be indicated on survey maps which may be posted near an RCA entrance or at other locations considered appropriate based on the radiation hazard and controls required.
- c. Radiation levels on survey maps shall be updated at a reasonable frequency. The data need not contain all survey results, but shall show radiation levels that indicate representative and specific significant radiation hazards.
6.3.10 Posting Requirements RP shall post areas within the Station to advise workers of radiological hazards. Area postings should provide additional instructions or information considered appropriate to make individuals aware of potential radiation exposures and to minimize exposures.
- a. Low Dose Waiting Areas (LDWA5)
Low Dose Waiting Areas are established to identify areas where dose rates are lower than other locations within the work area. When practicable, workers should be directed to remain in a LDWA unless they are actually needed at the work location. LDWAs should be posted: LOW DOSE WAITING AREA
- b. Radiation Area A Radiation Area is an area, accessible to individuals, in which radiation levels could result in an individual receiving a deep dose equivalent in excess of 0.005 rem in 1 hr (5 mrernlhr) at 30 centimeters from the source. These areas are posted:
CAUTION RADIATION AREA
- c. High Radiation Area A High Radiation Area is an area, accessible to individuals, in which radiation levels could result in an individual receiving a deep dose equivalent in excess of 0.1 rem in 1 hr (100 rnrem!hr) at 30 centimeters from the source. These areas are posted:
CAUTION, HIGH RADIATION AREA
DOMINION VPAP-2101 REVISION 34 PAGE 44 OF 110
- d. Locked High Radiation Area (LURA)
A LHRA is an area, accessible to individuals with radiation levels greater than or r equal to 1000 mrem per hour at 30 cm from the source of radiation or 30 cm from any surface that the radiation penetrates. These areas are posted: CAUTION, LOCKED HIGH RADIATION AREA
- e. Very High Radiation Area A Very High Radiation Area is an area, accessible to individuals, in which radiation levels could result in an individual receiving an absorbed dose in excess of 500 rad ml hr (500 radlhr) at 1 meter from the source. These areas are posted:
GRAVE DANGER, VERY HIGH RADIATION AREA
- f. Hot Spot A Hot Spot is a location where radiation levels near the source (e.g., within 2 to 3 cm) are significantly greater than radiation levels at 30 cm from the source. Hot Spots are identified so workers are aware of the higher radiation levels and can take precautions to minimize time spent near the location. Hot Spots are posted (the actual dose rate may by given after the wording but is not required):
CAUTION HOT SPOT (Dose Rate) NOTE: Health Physics personnel shall be exempt from the RWP issuance requirements during the performance of their assigned radiation protection duties, provided they comply with approved plant radiation protection procedures for entry into high radiation areas. [Commitment 3.2.5] 6.3.11 Access Controls for Posted High Radiation Areas
- a. A RWP authorizing access is required to enter Posted High Radiation Areas.
- b. Personnel permitted to enter Posted High Radiation Areas shall be provided with or accompanied by one or more of the following:
- 1. A survey instrument that continuously indicates the area dose rate.
DOMINION VPAP-2101 REVISION 34 PAGE 45 OF 110
- 2. A radiation monitoring instrument that continuously integrates the radiation dose rate in the area and alarms when a preset cumulative dose is received (e.g., DADs). Entry is permitted after the area dose rate is made known to personnel.
- 3. An HP Technician equipped with a survey meter. The HP Technician shall be responsible for performing radiological monitoring at the frequency specified on the RWP and for controlling activities within the area.
6.3.12 Access Controls for Posted Locked High Radiation Areas RP shall establish requirements to control access to Posted Locked High Radiation Areas. As a minimum, these controls shall include: [Commitment 3.2.5]
- a. Controls equivalent to those required for Posted High Radiation Areas shall be established.
- b. Locked door or doors and controls (e.g., locking, guarding) shall be established such that unauthorized personnel entry does not occur. [Commitment 3.2.4]
- c. Barricades established for the purpose of preventing entry to Locked High Radiation Areas (LHRA) shall be constructed of a substantial material (e.g., chain link mesh, shield plugs, etc.).
- d. Physical barriers used to control access should provide reasonable assurance that the barriers secure the area against unauthorized access and cannot be easily circumvented (i.e., individuals who incorrectly assume, for whatever reason, that they are authorized to enter the area, would be unlikely to disregard and/or circumvent the barrier).
- e. Keys to LHRA shall be maintained under the administrative control of HP Supervision or other designated HP personnel.
- f. Keys to LHRA shall only be issued to HP Technicians or an authorized HP individual who is trained and qualified in LHRA control procedures.
DOMINION VPAP-2l01 REVISION 34 PAGE 46 OF 110
- g. If the door is not equipped with a self-locking keyless-egress lock, then physical access control to the area during those periods when the access is unlocked shall be provided by one of the following: [Commitment 3.2.4]
- A Health Physics Technician
- An authorized HP individual who is trained and qualified in LHRA control procedures
- A qualified Access Control Guard
- h. If an individual will be working in a whole body dose rate field in excess of 1,000 mrem/hr, then in addition to requirements for entering a LHRA, additional controls will be established. As a minimum, these controls shall include:
- RWP briefings shall only be conducted by RP supervision. [Commitment 3.2.1]
radiological and operating conditions for entry shall be determined to
- f. beAcceptable stable for the duration of the access.
- Continuous coverage must be provided by an HP Technician to verify acceptable radiological conditions during personnel access.
- Visual and audio communications shall be established directly between the HP Technician and the worker at the job site prior to beginning work.
[Commitment 3.2.1] Maximum allowable stay time for the individuals in the area will be established. [Commitment 3.2.7]
DOMINION VPAP-2101 REVISION 34 PAGE 47 OF 110
- i. If an individual will be working in a whole body dose rate field in excess of 10,000 hr, then in addition to requirements for working in a whole body dose rate 1
mreni field in excess of 1,000 mremlhr, additional controls will be established. All work in the fuel transfer canal shall be subject to these controls. As a minimum, these controls shall include: [Commitment 3.2.1]
- RWP approval by the Manager Radiological Protection and Chemistry.
- A detailed survey of the job site shall be performed specifically evaluating the workers positions and orientation in the radiation field prior to issuing RWP for work. [Commitment 3.2.1]
- Facility Safety Review Committee (FSRC) approved procedures shall be used for all work activities. [Commitment 3.2.1]
- Requiring the issuance of a stop work order if dose rates are 1.5 times greater than those specified in the RWP under which the work is being performed.
[Commitment 3.2.1]
- j. Written approval of the Manager Radiological Protection and Chemistry for entry into areas where dose rates exist in excess of 100 remlhr. [Commitment 3.2.7]
6.3.13 Access Controls for Posted Very High Radiation Areas RP shall establish controls for access to Posted Very High Radiation Areas. As a minimum, the controls shall include the following: [Commitment 3.2.51
- a. Controls equivalent to those required for Posted Locked High Radiation Areas.
- b. If locks are used to restrict access, the locks shall be keyed so only keys designated as Very High Radiation Areas can open the locks.
- c. A specific (Special) RWP and associated documentation package:
[Commitment 3.2.7]
- 1. Shall be reviewed by FSRC.
- 2. Shall receive written approval of the Manager Radiological Protection and Chemistry and Plant Manager Nuclear.
- 3. Notifications to the Operations Shift Manager shall be made before entry into a Very High Radiation Area for under vessel/incore guide tube area entry.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 68. G2.3.4 068/BANKJNAPS/L/4/3.2/3.7//
Which ONE of the following provides the correct value for the associated Federal Annual Exposure Limits? Extremities SIdn Lens of eye A. 15 rem/yr 50 rem/yr 5 rem/yr B. 50 rem/yr 50 rem/yr 15 rem/yr C. 50 rem/yr 15 rem/yr 5 rem/yr D. 50 rem/yr 15 rem/yr 15 rem/yr
- a. Incorrect. Plausible since since these are all limits, but the numbers are different.
- b. Correct. These are the current limits found in 10CFR and VPAP-2101.
- c. Incorrect. Plausible since since these are all limits, but the numbers are different.
- d. Correct. Plausible since since these are all limits, but the numbers are different.
Radiation Control Knowledge of radiation exposure limits under normal or emergency conditions. (CFR: 41.12 / 43.4/45.10) Tier: 3 Technical
Reference:
VPAP-2101 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info: considered bank since although question may be somewhat different in format all of these are limits candidates are responsible for knowing.
DOMINION VPAP-2 101 REVISION 34 PAGE 31 OF HO 6.3.2 Regulatory Dose Limits and Controls Limits provided in this Step are 10 CFR 20 limits. Exceeding these limits is a violation of 10 CFR 20 and shall be appropriately addressed and reported. NOTE: Total Effective Dose Equivalent (TEDE) means the sum of deep-dose equivalent (from external exposures) and committed effective dose equivalent (from internal exposures). Deep-dose equivalent may be referred to as external whole body dose.
- a. Dose Limits for Radiation Workers Dose limits for occupationally exposed adult radiation workers, including occupational dose received at all facilities are:
Type Worker Dose Limit Total Effective Dose Equivalent (TEDE) 5 rein per calendar year Lens of Eye (lens dose equivalent) 15 rem per calendar year Skin (shallow dose equivalent) 50 rem per calendar year Extremities (shallow dose equivalent) 50 rem per calendar year
- b. Dose Limits to Embryo/Fetus of Declared Pregnant Woman If a woman is a Declared Pregnant Woman and has provided an estimated date of conception, then dose limits, which include occupational dose received at all 0 facilities and occupational dose received in the Controlled Area, are:
Os Type Dose Limit Dose equivalent to embryo/ferns of Declared 0.5 rem Pregnant Woman during entire pregnancy If, when pregnancy declared, dose is 0.450 rem or 0.05 rem for the remainder greater to the embryo/fetus, then allowable of the pregnancy remaining dose
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 69. G2.4.27 069/NEW//H/4/3.4/3.9//
In accordance with 0-FCA-1, Control Room Fire, RCS cooldown rate shall be and a 29-hour hold is NOT required prior to depressurizing the RCS provided___________ A. less than 15°F/hr; at least one CRDM fan is running. B. less than 15°F/hr; three CRDM fans are running. C. less than 50°F/hr; at least one CRDM fan is running. D. less than 50°F/hr; three CRDM fans are running.
- a. Incorrect. First part is correct. Second part is incorrect but plausible, one fan will provide head cooling but the supporting analysis is done based on three in operation, so the procedure only considers three fans. Having less than three running is for all intents and purposes having zero running.
- b. Correct. First part is correct. Second part is also correct. As discussed above three must be in operation to take any credit for their contribution to head cooling.
- c. Incorrect. First part is in correct but plausible since this is the admin limit, and the candidate who lacks detailed knowledge of the FCAs would likely default to this distractor. Second part incorrect but plausible as discussed in Distractor a.
- d. Incorrect. First part is incorrect but plausible as discussed in Distractor c. Second part is correct as explained above.
Emergency Procedures / Plan Knowledge of fire in the plant procedures. (CFR: 41.10 / 43.5 / 45.13) Tier: 3 Technical
Reference:
0-FCA-0.1 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
NUMBER PROCEDURE TITLE REVISION 36 O-FCA-1 CONTROL ROOM FIRE PAGE 37 of 70 -j_STEP_[f ACTION/EXPECTED RESPONSE I RESPONSE NOTOBTAINED I CAUTION:
- Changes to feedwater flow and steam flow should be made slowly.
Limit RCS cooldown to less than 15°F/hr. 31._ ESTABLISH RCSCOOLDOWN USING LV O SG PORVs TO LESS THAN 550°F U cQ- cs V D a) Check all SG PORVs AVAILABLE
- a) Periorm the following:
D 1) Locally open 1-MS-20, Decay Heat Release Valve Isolation. D 2) Use 1-MS-HCV-104, Decay Heat Release Valve, to dump steam to atmosphere. D IF Decay Heat Release Valve is NOT available, THEN send an Operator to MSVH to locally operate SG PORVs using ATTACHMENT 7, MAIN STEAM VALVE HOUSE OPERATIONS UNIT - D b) Cool down RCS by dumping steam to atmosphere D c) Maintain SG level greater than or equal to WR level corresponding to 33% Narrow Range level using ATTACHMENT 9, WIDE RANGE SG LEVEL CORRESPONDING TO 33% NARROW RANGE LEVEL (STEP 31 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 36 O-FCA-1 CONTROL ROOM FIRE PAGE 39 of 70 -_STEP ACTION/EXPECTEDRESPONSE j RESPONSENOTOBTAINED 33._ BLOCK SI INPUTS: D a) Obtain the Solid State Protection Panels, ESF Output Panels key from Appendix R Key Locker D b) Request Emergency Manager to direct D b) Have Operations personnel remove support personnel to remove ESF Output ESF Output fuses on both units. fuses on both units
- 34. DEPRESSURIZE RCS TO 1950 PSIG D Allow RCS pressure to decay to 1950 psig.
- 35. CONTINUE RCS COOLDOWN WHILE MAINTAINING THE FOLLOWING RCS CONDITIONS D a) PRZR Level 20%-29%
- D a) Adjust Charging flow.
b) RCS Cooldown Rate LESS THAN
- D b) Reduce steam dump flow.
15°F/HR D c) RCS Temperature and Pressure WITHIN
- c) Perform the following, as required:
ACCEPTABLE OPERATING REGION OF THE 15°F/HR COOLDOWN CURVE IN
- Adjust PRZR Heaters.
ATTACHMENT 12, REMOTE D
- Adjust steam dump flow.
MONITORING OPERATIONS D d) RCS Hot Leg Temperatures at Remote Monitoring Panel DECREASING NOTE: Steam Tables are located in each Auxiliary Shutdown Panel to aid in determining RCS Subcooling.
- 36. ESTABLISH AND MAINTAIN RCS Perform the following, as applicable:
SUBCOOLING GREATER THAN 175°F D
- Cool down RCS to achieve subcooling greater than 175°F.
- Energize PRZR Heaters to increase RCS pressure.
NUMBER PROCEDURE TITLE REVISION 36 O-FCA-1 CONTROL ROOM FIRE PAGE 67 of 70 -EE1-{ ACTION! EXPECTED RESPONSE I I RESPONSE NOT OBTAINED
- 55. COOL DOWN RCS OF THE APPLICABLE UNIT TO 190°F AND MAINTAIN RCS COOLDOWN RATE LESS THAN 15°F/HR (Continued) b) Adjust the following valves as necessary:
Unit 1: D
- 1-RH-FCV-1605, Residual Ht Rmvl Ht Exchs Recirc HDR Flow Cont Vv D
- 1-RH-HCV-1758, Residual Heat Removal Hx Outlet Flow Control Valve
- Unit 2:
D
- 2-RH-FCV-2605, Residual Ht Rmvl Ht Exchs Recirc HDR Flow Cont Vv
- 2-RH-FCV-2758, Residual Heat Removal Hx Outlet Flow Control Valve
- 56. CHECK IF 29 HOURS RCS HOLD - NOT REQU IRED D a) Verify three CRDM Fans - RUNNING a) GOTO Step 57.
ZOEEE) pcs
NUMBER PROCEDURE TITLE REVISION 36 O-FCA-1 CONTROL ROOM FIRE PAGE 68 of 70 H_STEP_fI ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED
- 57. HOLD RCS CONDITIONS FOR AT LEAST 29 HOURS:
a) Record Time/Date 29 hour hold began: Unit 1: Time: Date: D Unit 2: Time: Date: b) Maintain the following RCS conditions:
- 1) RCS temperature - LESS THAN 1) Perform the following for the 200°F applicable Unit:
- a. Locally adjust one or both of the following valves:
Unit 1: D
- 1-CC-MOV-100A D
- 1 -CC-MOV-1 OOB Unit 2:
- 2-CC-MOV-200A
- 2-CC-MOV-200B (STEP 57 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 36 O-FCA-1 CONTROL ROOM FIRE PAGE 690170 J_STEP ACTION/EXPECTEDRESPONSE j RESPONSENOTOBTAINED I
- 57. HOLD RCS CONDITIONS FOR AT LEAST 29 HOURS: (Continued)
- b. Adjust the following valves as necessary:
Unit 1:
- 1-RH-FCV-1605 D
- 1-RH-HCV-1758 Unit 2:
- 2-RH-FCV-2605
- 2-RH-HCV-2758 D 2) RCS pressure - 300 PSIG - D 2) Cycle PRZR Heaters and/or Auxiliary 350 PSIG Spray.
0 3) PRZR level - 20% - 29% 3) Adjust PRZR level using: 0
- Charging flow 0
- Seal Injection flow (if in-service) 0
- Letdown flow (if in-service)
- 4) SECURE SG SECONDARY:
0 a. WHEN Feedwater is no longer required, THEN remove Feedwater and/or AFW from service 0 b. Slowly depressurize all SGs by opening SG PORVs
Z 0 C L1j CO (1)10 0 5(0 <0 a_ LU 00 H z 2D o F 1O D 0 0 o ci)
- H (I)
G) D 0
- D I LU Z oH D 00 I-LU LU LU
-J a: U DOD I-H LU 0 0 C D a: LU QU) Z C -J C H LU LU LU LL UJ 0 N -J U U-H 0 a: 0 0 H a: 0
<H HU) a: Z D 0 Z a 0 (I) 0 0 (I) Z HZ LU a:
a: C D a- -J H
-Jo LU 0 C I >-
(1) Ow 0 U) 0 07 Z a: a: D a: LUU) HZ a: D 10 HJ LU 0 ILU I I LU H DH
- 0) (f)<
0 C\l U)LU -J
>- >- LU U U a:a: LU a: a: QN ZH a: LU w LUl OLU LU -
> > Ca ca -
0 DL Z Q d 10 10 (0 Co
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 70. G2.4.4 070/NEW//L/4/4.5/4.7//
Which ONE of the following would result in entry into the emergency plan? A. RCS identified leakage is determined to be 8.3 gpm. B. Unit 1 enters TS 3.0.3. C. An unconfirmed security threat is received. D. Offsite power was lost one hour ago and has not been restored.
- a. Incorrect. Plausible since the adoption on NEI EALs is fairly recent and the candidate who relies on past knowledge would erroneously select this distactor.
- b. Incorrect. Plausible since this is an EAL but only if the time to reach the mode or operating condition isnt met.
- c. Incorrect. Plausible since this is an EAL, but only for threats that have confirmation, however the candidate might erroneously believe that any threat would be an NOUE and an actual confirmed one could be an alert.
- d. Correct. Based on being lost for> 15 minutes this condition meets the NOUE threshold.
Emergency Procedures / Plan Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures. (CFR: 41.10/43.2/45.6) Tier: 3 Technical
Reference:
NAPS EAL maitrix, VPAP-2802 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
North Anna Power Station Revision I Emergency Action Level Technical Bases Document Attachment I Emergency Action Level Technical Bases SUI..1 Category: S System Malfunction Sub-category: 1 Loss of Power Initiating Condition: Loss of all offsite power to emergency busses for greater than 15 minutes EAL: SUtI Notification of Unusual Event Loss of all offsite AC power to Unit I (Unit 2)4160-Volt emergency busses H and J for> 15 mm. (Note 3) Note 3: The SEM should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time. Mode Applicability: I - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Basis: Prolonged loss of all offsite AC power reduces required redundancy and potentially degrades the level of safety of the plant by rendering the plant more vulnerable to a complete loss of AC power (station blackout). Unit I (Unit 2) 4160-Volt emergency busses H and J are the essential busses (ref. 1). The main generators are connected to the plant through the station service transformers (SSTs), which step the generator voltage down for distribution to the plant auxiliary systems. The generators are connected to the switchyard through the main transformers (MTs). A breaker on the output of Unit I generator allows the generator to be electrically disconnected from the SSTs and MTs; the Unit 2 generator does not have a generator breaker. When a unit is shut down the plant auxiliary systems are provided with electrical power from the switchyard through the MTs and SSTs or Reserve Station Service Transformers (RSSTs). Page 209 of 305
North Anna Power Station Revision I Emergency Action Level Technical Bases Document Attachment I Emergency Action Level Technical Bases SU6.1 Category: S System Malfunction Sub-category: 6 RCS Leakage ( Initiating Condition: RCS leakage EAL
)r e
yoç SU6.1 Notification of Unusual Event f 1 ( Unidentified or pressure boundary leakage> 10 gpm OR Identified leakage > 25 gpm Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Basis: The conditions of this EAL may be a precursor of more serious conditions and, as a result, is considered to be a potential degradation of the level of safety of the plant. Manual or computer-calculated water balance inventory methods are normally used to determine RCS leakage. Identified leakage is defined in Technical Specifications (ref. 2) as:
- Leakage from pump seals or valve packing (except reactor coolant pump seal water injection or leakoff) that is captured and conducted to collection systems or a sump or collecting tank.
- Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary leakage.
- RCS leakage through a steam generator to the Secondary System.
Page 257 of 305
North Anna Power Station Revision I Emergency Action Level Technical Bases Document Attachment I Emergency Action Level Technical Bases HU4.1 Category: H Hazards Sub-category: 4 Security Initiating Condition: Confirmed security condition or threat which indicates a potential
/ degradation in the level of safety of the plant EAL:
HU4.I Notifition of Unusual Event A security condition thtioes not involve a hostile action as reported by the Security Shift Supervisor ) fcCrJ &v o r 9) C 0L v A credible site-specific security threat notification OL O0O VL At_ A validated notification from NRC providing information of an aircraft threat Mode Applicability: All Basis: Security conditions which do not represent a potential degradation in the level of safety of the plant are reported under 10 CFR 73.71 or in some cases under 10 CFR 50.72. Security events assessed as hostile actions are classifiable under HA4.1, HS4.1 and HG4.1. A higher initial classification could be made based upon the nature and timing of the security threat and potential consequences. Consideration should be given to upgrading the emergency response status and emergency classification level in accordance with Dominions Security Plan, Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program. Page 186 of 305
North Anna Power Station Revision I Emergency Action Level Technical Bases Document Attachment I Emergency Action Level Technical Bases SU31 Category: S System Malfunction Sub-category: 3 Inability to Reach or Maintain Shutdown Conditions Initiating Condition: Inability to reach required shutdown within Technical Specification limits EAL: CYSfO 5U3.I Notification of Unusual Event Plant is not brought to required operating mode within Technical Specifications LCO action statement time Mode Applicability: I - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Basis: Limiting Conditions of Operation (LCOs) require the plant to be brought to a prescribed shutdown mode when the Technical Specification configuration cannot be restored. Depending on the circumstances, this may or may not be an emergency or precursor to a more severe condition. In any case, the initiation of plant shutdown required by the Technical Specification requires a four-hour report under 10 CFR 50.72 (b) non-emergency events. The plant is within its safety envelope when being shut down within the allowable action statement time in the Technical Specifications. An immediate declaration of an Unusual Event is required when the plant is not brought to the required operating mode within the allowable action statement time in the Technical Specifications. Declaration of an Notification of Unusual Event is based on the time at which the LCO specified action completion period elapses under Technical Specifications and is not related to how long a condition may have existed. Other Technical Specification shutdowns that involve precursors to more serious events are addressed by other EAL5. Page 231 of 305
DOMINION VPAP-2802 REVISION 33 PAGE 82 OF 199 NOTE: If a unit enters a limiting condition for operation (LCO) and a unit shutdown is started due to the LCO, the event is reportable even if shutdown is not completed. LCOs terminated by a unit shutdown for an unrelated reason are still reportable if the condition would not have been corrected within the LCO time limit for shutdown.
- 1. Initiation of plant shutdown (reduction of power or temperature) required by Technical Specifications. The initiation of plant shutdown does not include mode changes required by Technical Specifications if initiated after the plant is already in a shutdown condition. See EPIP-1.O1. [10 CFR 5O.72(b)(2)(i),
10 CFR 50.36(d)(1)(i)(A), 10 CFR 50.36 (d)(2)(i), NUREG 1022 Item 3.2.1]
- 2. Any event that results or should have resulted in ECCS discharge into the RCS as a result of a valid signal except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. tio CFR 50.72(b)(2)(iv)(A)]
- 3. Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when actuation results from and is part of a pre-planned sequence during testing or reactor operation.
[10 CFR 50.72(b)(2)(iv)(B)] NOTE: Notification to other government agencies has been or will be made is not necessarily an automatic notification to the NRC. Refer to NIJREG 1022, Event Reporting Guidelines 10 CFR 50.72 and 50.73, for discussions and examples or contact Station Licensing if clarification is needed. [NUREG-1022, Section 3.2.12]
- 4. Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned, or notification to other government agencies has been or will be made.
Such an event may include an onsite fatality or inadvertent release of radioactively contaminated materials. [Commitment 3.2.12] [10 CFR 50.72(b)(2)(xi)]
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 71. WEO2EK2. 1 07 IINEW//H/3/3 .4/3.9/4/
Given the following conditions:
- Unit 1 tripped from 100% power due to a spurious safety injection.
- The crew has transitioned to 1-ES-li, SI Termination.
The OATC is performing step 1, Reset Both Trains of SI, and notes that annunciator 1P-H2, AUTO SI BLKD TRNS A&B, is FLASHING. Which ONE of the following identifies the SSPS status, AND includes the action the crew should take? A. BOTH trains of SSPS have failed to RESET; reset both trains by closing and then opening Reactor Trip Breakers to clear the seal-in. B. BOTH trains of SSPS have failed to RESET; place both trains of SSPS in TEST and locally reset BOTH trains. C. ONE train of SSPS has failed to RESET; reset the affected train by closing and then opening Reactor Trip Breakers to clear the seal-in. D. ONE train of SSPS has failed to RESET; place the affected train of SSPS in TEST and locally reset the affected train.
- a. Incorrect. Plausible since the candidate who lacks detailed systems knowledge may conclude the reason for the status as being indicative of reset signal present but master relays not reset. Second part is plausible since this is one of the actions that would normally be performed as part of recovery actions anytime SI actuates, and again the candidate who lacks detailed systems knowledge may erroneously assume that this will remedy the situation.
- b. Incorrect. First part incorrect but plausible as noted above. Second part is correct as far as method, however the procedure (0-AP-0) precludes placing both trains in test, but again the person who has only cursory knowledge may conclude that this is the only method to address the given situation.
- c. Incorrect. First part is correct as the flashing annunciator indicates a train disagreement (i.e. one reset, and one NOT reset). Second part incorrect but plausible as discussed in Distractor a.
- d. Correct. First part is correct as discussed in Distractor c. Second part is also correct lAW 0-AP-0 (step 2 for train A, step 3 for train B).
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 SI Termination Knowledge of the interrelations between the (SI Termination) and the following: (CFR: 41.7 I 45.7) Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Tier: 1 Group: 2 Technical
Reference:
1-AP-O Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
C DeHtI H H1Ot NORTH ANNA POWER STATION ABNORMAL PROCEDURE NUMBER PROCEDURE TITLE REVISION 2 1 -AP-O RESETTING SI LOCALLY (WITH THREE ATTACHMENTS) PAGE 1 of 9 PURPOSE To provide instructions for responding to a failure of a train of Safety Injection to reset from the Control Room. ENTRY CONDITIONS This procedure is entered if a train of Safety Injection will not reset from the Control Room. The following are symptoms that a train of SI has not reset: Failure of a train of Solid State Protection based on equipment not repositioning as required
- Annunciator Panel P-H2, AUTO SI BLKD TRNS A&B - NOT LIT or FLASHING 4 _c CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 2 1 -AP-O RESETTING SI LOCALLY PAGE 2 of 9 -Ei--F ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
ç d j ( C -1 Reactor Trip Breakers MUST NOT be reclosed to prevent SI rein itiation.
Only ONE train of Solid State Protection shall be placed in test. NOTE:
- Component locations, relay layout and relay positions are provided on Attachment 3, SOLID STATE PROTECTION FIGURES.
- If available, then an Instrument Technician should assist in Solid State Protection relay reset.
- Equipment on the affected train will not shutdown remotely until SI is reset on the affected train.
- Attachment 2, Table 1 RELAY FUNCTIONS provides information on relays and affected equipment.
- 1. DETERMINE FAULTED TRAIN BY OBSERVING PLANT STATUS:
D a) TRAIN A-RESET a) GO TO Step 2 to reset Train A. 11 b) TRAIN B - RESET b) GO TO Step 3 to reset Train B. J c) RETURN TO Procedure And Step In Effect
NUMBER PROCEDURE TITLE REVISION 2 1-AP-O RESETTING SI LOCALLY PAGE 3 of 9 j_STEP ACTION/EXPECTEDRESPONSE I I RESPONSENOTOBTAINED I 2._ RESET TRAIN A LOCALLY: D a) Obtain Solid State Protection keys and card puller from Control Room C b) Have TWO individuals proceed to UNIT 1 Rack Room c) Open the following Cabinets: C
- 1-EI-CB-62A, SAFEGUARDS TEST CABINET TRAIN A C
- 1-EI-CB-47E, SOLID STATE PROT OUTPUT CABINET TRAIN A d) In 1-EI-CB-47E do the following:
C 1) PIaceMODESELECTORSWITCH in the TEST position
- 2) SIMULTANEOUSLY depress and HOLD the following pushbuttons while performing Step 2.e: LI C
- A SLAVE TEST Yellow pushbutton
\
C
- BSLAVE TEST Blue pushbutton (STEP 2 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 2 1-AP-O RESETTING SI LOCALLY PAGE 4 of 9 -_STEP FI ACTION/EXPECTED RESPONSE 1 RESPONSE NOT OBTAINED JL5
- 2. RESET TRAIN A LOCALLY: (Continued) e) In 1-EI-CB-62A have second individual do the following:
J 1) Place the S821 RESET switch in the RESET position and release D 2) Repeat the reset by placing S821 RESET switch in the RESET position and release f) In 1 -El-CB-47E release the following pushbuttons: D
- A SLAVE TEST Yellow pushbutton
- B SLAVE TEST Blue pushbutton D g) Have Unit 1 SRO check if I g) Have Unit 1 SRO determine if local 1-SI-MOV-1867A, BIT Inlet Isolation Valve closure of BIT Inlet Valve, is required.
can be closed from Control Room IF local closure is required, THEN do the following: I 1) Open 1-EE-BKR-1H1-2N Dl, BIT Inlet Isolation Valve Circuit Breaker, 1-Sl-MOV-1867A in the Unit 1 Cable Vault. D 2) Notify Unit 1 SRO that 1 -Sl-MOV-1 867A can be locally closed. (STEP 2 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 2 1 -AP-O RESETTING SI LOCALLY PAGE 5 of 9 f-_jACTION!EXPECTED RESPONSE RESPONSE NOT OBTAINED H_STEP
- 2. RESET TRAIN A LOCALLY: (Continued)
E1 h) Have Unit 1 SRO check if h) Have Unit 1 SRO determine if local 1-CH-MOV-1289A, Normal Charging opening of Normal Charging Isolation Isolation Valve can be opened from Control Valve, is required. Room IF local opening is required, THEN do the following: D 1) Open 1-EE-BKR-1H1-2N G1, Normal Charging Isolation Valve Circuit Breaker, 1-CH-MOV-1289A, in the Unit 1 Cable Vault. EJ 2) Notify Unit 1 SRO that 1-CH-MOV-1289A can be locally opened. 1J i) GOTOSTEP4
NUMBER PROCEDURE TITLE REVISION 2 1 -AP-O RESETTING SI LOCALLY PAGE 6 of 9 j_STEP ACTION/EXPECTEDRESPONSE I I RESPONSENOTOBTAINED I
- 3. RESET TRAIN B LOCALLY:
D a) Obtain Solid State Protection keys and card puller from Control Room b) Have TWO individuals proceed to UNIT 1 Rack Room c) Open the following Cabinets: D
- 1-EI-CB-62B, SAFEGUARDS TEST CABINET TRAIN B
- 1-EI-CB-47F, SOLID STATE PROT OUTPUT CABINET TRAIN B d) In 1 -EI-CB-47F do the following:
D 1) Place MODE SELECTOR SWITCH in the TEST position
- 2) SIMULTANEOUSLY depress and HOLD the following pushbuttons while performing Step 3.e:
D
- ASLAVE TEST Yellow pushbutton D
- B SLAVE TEST Blue pushbutton (STEP 3 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 2 1 -AP-O RESETTING SI LOCALLY PAGE 7 of 9 H_STEP ACTION/EXPECTED RESPONSE I RESPONSE NOTOBTAINED
- 3. RESET TRAIN B LOCALLY: (Continued) e) In 1-EI-CB-62B have second individual do the following:
C 1) Place the S821 RESET switch in the RESET position and release C 2) Repeat the reset by placing S821 RESET switch in the RESET position and release f) In 1-EI-CB-47F release the following pushbuttons: C
- A SLAVE TEST Yellow pushbutton C
- B SLAVE TEST Blue pushbutton C g) Have Unit 1 SRO check if C g) Have Unit 1 SRO determine if local 1-SI-MOV-1867B, BIT Inlet Isolation Valve closure of BIT Inlet Valve, is required.
can be closed from Control Room IF local closure is required, THEN do the following: C 1) Open 1-EE-BKR-1J1-2N C3, BIT Inlet Isolation Valve Circuit Breaker, 1-Sl-MOV-1867B in the Unit 1 Cable Vau It. C 2) Notify Unit 1 SRO that 1-SI-MOV-1867B can be locally closed. (STEP 3 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 2 1-AP-O RESETTING SI LOCALLY PAGE 8 of 9 H_STEP ACTION/EXPECTEDRESPONSE I RESPONSENOTOBTAINED I
- 3. RESET TRAIN B LOCALLY: (Continued)
EJ h) Have Unit 1 SRO check if D h) Have Unit 1 SRO determine if local 1-CH-MOV-1289B, Normal Charging opening of Normal Charging Isolation Isolation Valve can be opened from Control Valve, is required. Room IF local opening is required, THEN do the following:
- 1) Open 1-EE-BKR-1J1-2N H3, Normal Charging Isolation Valve Circuit Breaker, 1-CH-MOV-1289B, in the Unit 1 Cable Vault.
D 2) Notify Unit 1 SRO that 1-CH-MOV-1289B can be locally opened. NOTE: Lo-Lo Pressurizer Pressure SI and High Steam Line Flow SI may not be manually blocked when required by plant conditions. Auto SI will not re-initiate as long as P4 is not reset by cycling the Reactor Trip Breakers. NOTE: When Accumulator valves are closed and the SI relays are reset in this procedure, then the valves associated with the reset train will reopen, if not de-energized.
- 4. PERFORM ATTACHMENT 2, SI RELAY VERIFICATION
z 0 > 00) Lii I
-J LU -J
-J < 0 HO
-J LU I 0
> z o z LU UJ -Q 0
0 C p Cl) (15 0 ci LU I LU 0 H 0 0.. Cl) UJ (15 I 0) 0 C C 0 z Cl) U) 0 0 :3 wH 0 LU EW W I D Cl) 0 CJ C) ( (0 LU cS icr 5o u I 0 0 g 0 0 Ci o u, I I c3) øH WH WU) 15) 0 c
- 0 Q W 0.. c .4-LU Cl)j Cl) CI)- Cl)i COwf 0 0 H C - - - -o H
- i5 0 OC OQ 0 15) ZH 15)
. I Hu LU LU
..0 ILU co D D
z
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 72. WEO5EK3 .3 072/NEW//H/4/4 .0/4.1/4!
Given the following conditions:
- Unit 1 was initially at 100% power with i-FW-P-2 tagged out.
- A Reactor Trip occurred, and multiple failures have resulted in a loss of all AFW pumps.
- The crew has transitioned to 1-FR-Hi, Response to Loss of Secondary Heat Sink.
Attempts to start a MEW pump proved unsuccessful, and the crew is attempting to establish SG feed flow from the Condensate system. Which ONE of the following describes how conditions are established to feed from the Condensate system, includes the reason? A. Depressurize ONLY one SG to reduce the probability of a PTS condition. B. Depressurize ONLY one SG to maximize time before bleed and feed criteria are met. C. Depressurize ALL SGs to ensure even cooling of the RCS. D. Depressurize ALL SGs to minimize the probability of an undesired SI actuation.
- a. Incorrect. Plausible since less RCS cooldown is implied by only doing one loop, and for the reference plant the WOG discusses only doing one SG, but this is not the method and reason at NAPS.
- b. Incorrect. Plausible since only doing one implies inventory would remain longer in the remaining two but as noted above, this is not the method and reason at NAPS.
- c. Incorrect. Plausible since the candidate may erroneously assume that this is desired especially since RCPs are stopped because AFW could not be immediately restored; this is the bases for some actions in the EOPs that direct starting a cooldown and while this is an argument that could be made, it is not the reason for the deviation from the WOG guidance.
- d. Correct. This action will provide timely depressurization and minimize the probability of an undesired SI which would complicate recovery actions.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Loss of Secondary Heat Sink Knowledge of the reasons for the following responses as they apply to the (Loss of Secondary Heat Sink) (CFR: 41.5/41.10,45.6,45.13) Manipulation of controls required to obtain desired operating results during abnormal, and emergency situations. Tier: 1 Group: 1 Technical
Reference:
EOP FR-Hi, FR-H.1 background Document, NAPS Step Deviation document Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info: This is a Step deviation from the WOG Background document.
NUMBER PROCEDURE TITLE REVISION 20 1 -FR-H 1 RESPONSE TO LOSS OF SECONDARY HEAT SINK
. PAGE 16 of 41 j_STEP ACTION/EXPECTED RESPONSE I RESPONSE NOTOBTAINED NOTE: To prevent an undesired Main Steamline Isolation, each Main Steamline flow should be kept less than 1 .0E6 LBM/HR. -r j-,--
8._ INITIATE DEPRESSURIZATION OftAy,iSGs o TO LESS THAN 610 PSIG BY DUMPING STEAM TO CONDENSER AT MAXIMUM RATE: D a) Verify Condenser Steam Dumps - a) Manually or locally depressurize SGs AVAILABLE using: D *SGPORVs OR EJ
- Decay Heat Release Valve using ATTACHMENT 6, USING DECAY HEAT RELEASE VALVE FOR COOLDOWN.
D IF depressurization is initiated, THEN GO TO Step 9. D IF unable to depressurize SOs, THEN GOTO Step 13. (STEP 8 CONTINUED ON NEXT PAGE) cS s v?CLc
NUMBER PROCEDURE TITLE REVISION 20 1 -FR-H 1 RESPONSE TO LOSS OF SECONDARY HEAT SINK
. PAGE 17 of 41
-j_STEP_ ACTION/EXPECTEDRESPONSE I RESPONSENOTOBTAINED I
- 8. INITIATE DEPRESSURIZATION OF ALL SGs TO LESS THAN 610 psig BY DUMPING STEAM TO CONDENSER AT MAXIMUM RATE: (Continued) b) Transfer Condenser Steam Dump to Steam Pressure Mode:
EJ 1) Put both Steam Dump Interlock switches to OFF/RESET D 2) Put Steam Dump controller to MANUAL D 3) Put Mode Selector switch to STEAM PRESS D 4) Verify or reduce Steam Dump demand to zero D 5) Put both Interlock switches to ON
- 6) Check Panel P-F3, P-12 PERM D 6) Momentarily place both Steam Dump TAVG <543F BLOCK SI BYP STM Intlk Switches in BYP INTK and verify DUMPS NOT LIT
- Panel P-F4, STM DUMP COOLDOWN VLV AVAILABLE LIT. -
D c) Raise Steam Dump Controller demand to dump steam
NUMBER PROCEDURE TITLE REVISION 20 1 -FR-H 1 RESPONSE TO LOSS OF SECONDARY HEAT SINK
. PAGE 18 of 41 H_STEP_fj ACTION/EXPECTEDRESPONSE I RESPONSENOTOBTAINED CAUTION: After blocking automatic SI actuation, manual SI actuation may be required if conditions degrade.
9._ CHECK HIGH STEAM FLOW SI BLOCK AND CONDENSER STEAM DUMP COOLDOWN INTERLOCK CONDITIONS: D a) Check Panel P-F3, P-12 PERM TAVG a) Do the following:
<543F BLOCK SI BYP STM DUMPS LIT -
C
- WHEN Tave is less than 543°F, THEN perform Step 9.
OR C
- Hold both Steam Dump Intlk Switches in BYP INTK until Tave is less than 543°F, then perform Step 9.
C Continue with Step 10. C b) Momentarily place both High Steam Flow SI Block Switches in BLOCK C c) Verify Panel P-G3, STM LINE PRESS-FLOW SI BLKD TRNS A&B LIT - C d) Verify Panel P-F4, STM DUMP C d) Momentarily place both Steam Dump COOLDOWN VLV AVAILABLE LIT - Intik Switches in BYP INTK. NOTE: Condensate Pump Recirc Valve adjustment or Feed Pump Recirc Valve closure may be required to increase Feed flow.
- 10. CHECK AT LEAST ONE SG PRESSURE - C Continue dumping steam AND RETURN TO LESS THAN 610 PSIG Step 9.
NUMBER PROCEDURE TITLE REVISION 20 1 -FR-H 1 RESPONSE TO LOSS OF SECONDARY HEAT SINK
. PAGE 19 of 41 H_STEP_fI ACTION/EXPECTED RESPONSE I RESPONSE NOTOBTAINED I
- 11. CHECK IF DEPRESSURIZATION OF SGs SHOULD BE STOPPED:
a) Verify feed flow to at least one SG - a) Continue to depressurize SGs at RESTORED AS INDICATED BY ANY OF maximum controllable rate. WHEN one of THE FOLLOWING: the following conditions is met, THEN GOTO Step lib: D
- Total AFW flow to SG(s) GREATER THAN 340 GPM
- BOTH of the following satisfied:
C
- SG Wide range level(s) -
INCREASING C
- Feed flow to at least one SG -
GREATER THAN 0.7E6 LBM/HR C
- Core exit TCs - DECREASING BOTH of the following satisfied: C
- Feed flow to at least one SG is greater than 0.7E6 LBM/HR C
- SG Wide range level(s) -
INCREASING OR C
- Core exit TCs - DECREASING C
- Narrow Range level in at least one SG greaterthan 11%[22%]
OR OR C
- Narrow Range level in at least one SG GREATER THAN 11% [22%J
- C
- Any SG pressure has decreased to 190 psig C b) Stop depressurizing all SGs
NUMBER PROCEDURE TITLE REVISION 20 1 -FR-H 1 RESPONSE TO LOSS OF SECONDARY HEAT SINK
. PAGE 20 of 41 J_STEP ACTION/EXPECTEDRESPONSE I RESPONSENOTOBTAINED I
NOTE: If this procedure is exited in Step 12b before narrow range level in at least one SG is greater than 11% [22%], then re-entry into 1-FR-H.1 is not required unless adequate feed flow is lost, as indicated by Core Exit TCs or SG wide range level response. Feed flow may be adjusted to control plant cooldown rate and PRZR level decrease, as long as SG wide range level(s) continue to increase and Core Exit TCs continue to decrease. CHECK SG LEVELS: D a) Narrow range level in at least one SG - D a) Verify adequate feed flow, as indicated by GREATER THAN 11% [22%] Core Exit TCs decreasing and SG wide range level(s) increasing. IF adequate feed flow to at least one SG is verified, THEN maintain feed flow to restore narrow range level greater than 11% [22%] and GOTO Step 12b. D IF adequate feed flow is NOT verified, THEN GO TO Step 13. C b) RETURN TO procedure and step in effect CHECK SG WIDE RANGE LEVEL IN C RETURN TO Step 1. ANY TWO SGs LESS THAN 14% [32%]
STEP DESCRIPTION TABLE FOR FR-H.l Step 7 STEP: Try To Establish Feed Flow From Condensate System PURPOSE: To direct the operator in establishing condensate flow as an alternative (or supplement) to AFW and main FW flow BASIS: The condensate system is the next source of water readily available to the operator for use in reestablishing the secondary heat sink.
> s_,__ f\
In order to depressurize at least one SG to less than the shutoff head pressure of the condensate system pumps, the RCS must be depressurized below (A.06) psig to allow blocking of the low steamline pressure SI and low PRZR pressure SI signals. If these signals were allowed to actuate, feedline and steamline isolation actuation signals may have to be reset. Feedline isolation may still occur on a reactor trip signal coincident with the low Tavg signal. Auxiliary spray is used to depressurize the RCS, if letdown is in service, since it provides a maximum cooling to the primary system while allowing no loss of primary water inventory. Normal spray is not available since RCPs are stopped (Step 4). If letdown is not in service, PRZR PORVs are used to avoid thermal stresses to the auxiliary spray nozzles. However, if the PRZR PORVs cannot be used, auxiliary spray must be used. Depressurization of the SG(s) is accomplished through the condenser steam dump, PORV, or other means if required. Footnote (0.09) defines the steam generator pressure requirement that will allow the condensate pump to provide adequate feedwater flow for decay heat removal. Minimum condensate flow for condensate pump protection, which is provided by a recirculation line flow control valve, is typically much greater than the flow required to remove decay heat. Reducing SG pressure to the condensate pump discharge header pressure for recirculation would permit the condensate pumps to inject into the SG with adequate feed flow for decay heat removal. FR-H.1 Background 83 HP-Rev. 2, 4/30/2005 HFRH1BG .doc
STEP DESCRIPTION TABLE FOR FR-H.l Step 7 The optimum number of SGs to depressurize to less than 0.09 psig, in the case of the reference plant, is one because certain benefits are realized by depressurizing only one SG. The likelihood of reaching the criteria for initiation of RCS bleed-and-feed is reduced because only a single SG is steamed. Additionally, the accompanying reduction in pressurizer level and RCS subcooling is less severe, which in turn reduces the likelihood that manual SI actuation will be required based on degraded plant conditions. Thus, before the SG is depressurized it should be isolated from the other SGs. s ç 1eu ACTIONS: 5 5 5 o Determine if letdown is in service o Depressurize RCS to less than (A.06) psig using eiher auxiliary spray or one PRZR PORV o Block low steamline pressure SI and low PRZR pressure SI signals o Depressurize at least one SG to less than (0.09) psig by dumping steam usi ng:
- Steam dump to condenser - SG PORV - Other plant specific means o Establish condensate flow INSTRUMENTATION:
o RCS pressure indication o Letdown system valves position indication o Auxiliary spray valve position indication o PRZR PORV position indication o SI signal status indication o SG pressure indication o Condenser steam dump valves position indication o SG PORVs position indication o Condensate system valves position and pumps status indication FR-H.1 Background 84 HP-Rev. 2, 4/30/2005 HFRH1BG .doc
STEP DESCRIPTION TABLE FOR FR-H.1 Step 7 CONTROL/EQUIPMENT: o Switches for:
- Auxiliary spray valve - PRZR PORV - Condensate system valves and pumps - SI signal block - Condenser steam dump valves - SG PORVs o Plant specific controls to dump steam by other means KNOWLEDGE:
At least one SG should be depressurized to a pressure that allows the condensate pump to deliver flow at least equal to that of which is used for decay heat removal. Providing condensate pump flow equal to the minimum flow used for recirculation, satisfies the flow requirement. PLANT-SPECIFIC INFORMATION: o (A.O6) PRZR pressure 50 psi below permissive to block SI. o (0.09) Condensate pump discharge header pressure for minimum flow operation on recirculation, minus allowances for normal channel accuracy. o Means to establish condensate flow o Means to depressurize SGs FR-H.1 Background 85 HP-Rev. 2, 4/30/2005 HFRH1BG .doc
.JI?S £Gi NAPS EOP: TITLE: REVISION:
FR-H.1 Response to Loss of Secondary Heat Sink 01-12-2010 EOP Step ERG Step Sequence No. 8N1 Added note on limiting steam flow. The note was added to inform the operator that excessive steam flow may cause the Hi steam flow with low Tavg SI and steam line isolation logic to be satisfied. The operator is therefore warned to maintain steam flow below a value that can cause the steam line isolation. Steam line isolation would complicate recovery actions. No adverse setpoint is used in this caution because any errors associated with an adverse environment would be seen equally at the meter and at the SllSteamline isolation bistable. Since the purpose of the caution is to prevent bistable actuation, the adverse setpoint is not needed. 8 7 (7) The step for establishing condensate flow has been expanded into several separate high level steps. See above for list of major actions. This step, along with subsequent steps, depressurizes all SGs to below the CN pump shut off head to enable the CN pump to feed the generators. Changed guidance to depressurize all rather than at least one SG. Depressurizing all SGs at approximately the same rate will prevent the steam line delta pressure SI. Because feedline isolation accompanies the SI, it is undesirable to receive an SI signal at this point. The intent of depressurizing at least one SG is met. Added information on use of the steam dump valves. Provided explicit guidance on using steam dump valve to cooldown based on operator feedback ER-91-025. Added decay heat release valve to the RNO. The decay heat release valve was added because it is another plant specific method for dumping steam. This change is consistent with the intent of the ERG. 9C This caution informs the operator that after the high steam flow signal is blocked in the next step, the operator will be relied upon to initiate SI if it is required. 9 This step blocks the high steam flow with low Tavg SI. Since a feedwater isolation signal accompanies an SI, it is important to block the SI signal. This step is contingent upon RCS Tavg below the block permissive setpoint (nominally 543 °F) and is therefore continuous action. Page 8 of 23
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 73. WEll EA2. 1 073/BANK/NAPS: 605 84/H14/3 .4/4.2/4/
Following a large-break LOCA, the crew has completed 1-ES-i .3, Transfer to Cold Leg Recirculation, Step 9, Align SI System For Cold Leg Recirculation. The following conditions exist:
- Amps and flow are oscillating on both Low-head SI pumps.
- The STA reports that a RED path exists on heat sink.
Which ONE of the following describes the correct actions AND implementation of procedures for these conditions? A. Perform 1-ES-i .3, Attachment 3, Containment Sump Screen Blockage or Loss of Suction; Do NOT implement i-FR-Hi, Response to Loss of Secondary Heat Sink, until directed in 1-F-i, Loss of Reactor or Secondary Coolant. B. Perform i-ES-1.3, Attachment 3, Containment Sump Screen Blockage or Loss of Suction; Do NOT implement 1-FR-Hi, Response to Loss of Secondary Heat Sink, until directed in Attachment 3 OR i-ECA-i.i, Loss of Emergency Coolant Recirculation. C. Immediately transition to i-FR-H.1, Response to Loss of Secondary Heat Sink. D. Immediately transition to 1-ECA-i.1, Loss of Emergency Coolant Recirculation.
- a. Incorrect. Plausible since the normal return path is to i-E-i which has a transition to 1-ECA-i.1, the candidate who does not have detailed knowledge of the requirements of the procedures for the given conditions may select this distractor.
- b. Correct. Based on the conditions given there is indication of sump blockage; since step 9 is complete foldout item 1 applies and takes precedence over the procedure note which would allow implementation of FRPs after step 9.
- c. Incorrect. Plausible since this is the rule for FRP implementation. ES-i .3 has an exception to this rule, and also as noted above there is further exception based on plant conditions; as such the candidate who lacks detailed knowledge of the procedure may default to this distractor
- d. Correct. Plausible since this procedure may ultimately be implemented, but not directly until ES-i .3 corrective actions are initiated per Att. 3, again though, if the candidate lacks detailed knowledge of the attachment and what it does, they may select this distractor.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Loss of Emergency Coolant Recirculation Ability to determine and interpret the following as they apply to the (Loss of Emergency Coolant Recircu lation) (CFR: 43.5 /45.13) Facility conditions and selection of appropriate procedures during abnormal and emergency operations. Tier: 1 Group: 1 Technical
Reference:
EOP ES-1.3 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
NORTH ANNA POWER STATION EMERGENCY PROCEDURE NUMBER PROCEDURE TITLE REVISION 22 1 -ES-i .3 TRANSFER TO COLD LEG RECIRCULATION PAGE (WITH FOUR ATTACHMENTS) 1 of 10 PURPOSE To provide instructions for transferring the Safety Injection System to the Cold Leg Recirculation Mode. ENTRY CONDITIONS This procedure is entered from:
- 1 -E-1 LOSS OF REACTOR OR SECONDARY COOLANT,
- 1 -ECA-2.1, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS,
- 1-FR-C.1, RESPONSE TO INADEQUATE CORE COOLING,
- 1-FR-C.2, RESPONSE TO DEGRADED CORE COOLING,
- 1-FR-C3, RESPONSE TO SATURATED CORE COOLING,
- 1-FR-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK, or
- RWST level has decreased to 23% and Cold Leg Recirculation is required.
CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 22 1 -ES-i 3 TRANSFER TO COLD LEG RECIRCULATION PAGE 2 of 10 -j_STEP_i] ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED CAUTION:
- SI recirculation flow to the RCS should be maintained at all times.
Transfer to recirculation mode may cause high radiation levels in the Auxiliary Building. NOTE:
- Step 1 through Step 9 should be performed without delay. FRs should not be implemented prior to completion of Step 9.
- IF Sump Blockage occurs, THEN FRs should NOT be implemented until directed in ATTACHMENT 3, CONTAINMENT SUMP STRAINER BLOCKAGE OR LOSS OF SUCTION OR 1-ECA-1 .1, LOSS OF EMERGENCY COOLANT RECIRCULATION.
- 1. RESET BOTH TRAINS OF SI C Perform 1-AP-0, RESETTING SI LOCALLY, while continuing with this procedure.
2._ VERIFY RECIRC SPRAY PUMP STATUS - C Align and start at least two Recirc Spray AT LEAST TWO PUMPS ALIGNED AND pumps. RUNNING
- 3. VERIFY PROPER SERVICE WATER SYSTEM OPERATION:
C a) At least two Service Water Pumps - C a) Manually start pumps. RUNNING b) Verify Service Water Supply to CC Heat C b) Manually close valves. Exchangers CLOSED: C
- 1-SW-MOV-108A C
- 1-SW-MOV-108B (STEP 3 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 22 1 -ES-i 3 TRANSFER TO COLD LEG RECIRCULATION PAGE 3 of 10 -J_STEP_[j ACTION/EXPECTED RESPONSE RESPONSE NOTOBTAINED I
- 3. VERIFY PROPER SERVICE WATER SYSTEM OPERATION: (Continued) c) Verify Service Water Supply and Return C c) Manually open valves.
Header Recirc Spray Heat Exchanger Isolation Valves OPEN: C
- 1-SW-MOV-1O1A C
- 1-SW-MOV-1O1C C
- 1-SW-MOV-105A C
- 1-SW-MOV-105C C
- 1-SW-MOV-1O1B C
- 1-SW-MOV-1O1D C
- 1-SW-MOV-105B C
- 1-SW-MOV-105D (STEP 3 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 22 1 -ES-i 3 TRANSFER TO COLD LEG RECIRCULATION PAGE 4 of 10 -_STEP [j ACTION/EXPECTEDRESPONSE I RESPONSENOTOBTAINED
- 3. VERIFY PROPER SERVICE WATER SYSTEM OPERATION: (Continued) d) Verify Service Water Supply and Return D d) Manually open valves on running Recirc Header Recirc Spray Heat Exchanger Spray pumps.
Isolation Valves for RUNNING RECIRC SPRAY PUMPS OPEN: - 1 -RS-P-1 A: D
- 1-SW-MOV-103A D
- 1-SW-MOV-104A 1 -RS-P-2A:
EJ
- 1-SW-MOV-103D D
- 1-SW-MOV-104D 1 -RS-P-1 8:
C
- 1-SW-MOV-103B C
- 1 -SW-MOV-1 04B 1 -RS-P-2B:
C
- 1-SW-MOV-103C C
- 1-SW-MOV-104C
NUMBER PROCEDURE TITLE REVISION 22 TRANSFER TO COLD LEG RECIRCULATION 1 -ES-i 3. PAGE 5 of 10 j_STEP [] ACTION/EXPECTEDRESPONSE j RESPONSENOTOBTAINED I
- 4. CHECK CHARGING PUMP STATUS - D Manually start pumps.
TWO CHARGING PUMPS RUNNING
- 5. VERIFY LOW-HEAD SI PUMPS RUNNING - D Manually start pumps.
- 6. ALIGN SEAL INJECTION AND CHARGING PUMP RECIRC:
a) Close 1-CH-MOV-1 370 D a) Close 1-CH-HCV-1186. D Locally close 1 -CH-MOV-1 370, RCP Seal Water Injection Isolation Valve, as time permits. b) Verify Charging Pump Recirc Isolation D b) Manually align valves as necessary. Valves CLOSED D
- 1-CH-MOV-1275A 0
- 1-CH-MOV-1275B 0
- 1-CH-MOV-1275C
- 7. VERIFY RWST LEVEL LESS THAN 15%
- 0 Do NOT continue with this procedure until RWST Level is less than 15%.
O WHEN RWST Level is less than 15%, THEN continue with Step 8.
NUMBER PROCEDURE TITLE REVISION 22 1 -ES-i 3 TRANSFER TO COLD LEG RECIRCULATION PAGE 6 of 10 _STEP_fI ACTION/EXPECTED RESPONSE I RESPONSE NOTOBTAINED VERIFY POWER AVAILABLE FOR COLD LEG RECIRCULATION: a) A Low-Head SI Valves - ENERGIZED: D a) Stop i-SI-P-iA, A Low-Head SI Pump. D
- 1-SI-MOV-1860A D
- i-SI-MOV-1862A b) B Low-Head SI Valves - ENERGIZED: G b) Stop 1-SI-P-i B, B Low-Head SI Pump.
D
- 1-SI-MOV-1860B D
- i-SI-MOV-1862B CAUTION: Charging or Low-Head Pumps NOT aligned for Cold Leg Recirculation and taking suction from the RWST must be stopped when RWST level decreases to 8%.
9.. ALIGN SI SYSTEM FOR COLD LEG Manually align valves in sequence as RECIRCULATION: necessary. a) Verify Low-Head SI Pump - AUTO IF at least one flow path from the ALIGNMENT: Containment Sump to the RCS cannot be established OR maintained, THEN do the
- 1) Low-Head SI Pump Discharge following:
Valves to Charging Pumps - OPEN:
- IF 1-E-0, Attachment 3, PRIMARY
- 1-Sl-MOV-1863A PLANT VENTILATION ALIGNMENT is NOT completed, THEN initiate
- 1-SI-MOV-1863B ATTACHMENT 2, PRIMARY PLANT VENTILATION ALIGNMENT.
- GOTO1-ECA-1.1,LOSSOF EMERGENCY COOLANT RECIRCULATION, STEP 1.
(STEP 9 CONTINUED ON NEXT PAGE)
PROCEDURE TITLE REVISION 22 TRANSFER TO COLD LEG RECIRCULATION PAGE 7 of 10 ACTION/EXPECTED RESPONSE J I RESPONSE NOT OBTAINED
- 9. ALIGN SI SYSTEM FOR COLD LEG RECIRCULATION: (Continued)
- 2) Low-Head SI Pump Recirc Valves -
CLOSED:
- 1-SI-MOV-1885A D
- 1-Sl-MOV-1885C D
- 1-SI-MOV-1885B D
- 1-Sl-MOV-1885D
- 3) Low-Head SI Pump Suction From Containment Sump OPEN:
EJ
- 1-Sl-MOV-1860A D
- 1-Sl-MOV-1860B
- 4) Low-Head SI Pump Suction From RWST CLOSED:
LJ
- 1-SI-MOV-1862A D
- 1-Sl-MOV-1862B
- 10. ALIGN CHARGING PUMP SUCTION:
a) Close Charging Pump Suction From RWST Isolation Valves: D
- 1-CH-MOV-11158 D
- 1-CH-MOV-1115D b) Verify continued flow of Charging Pumps -
INDICATED: D
- 1-SI-FI-1943 D
- 1-SI-FI-1943-1
NUMBER PROCEDURE TITLE REVISION 22 1 -ES-i 3 TRANSFER TO COLD LEG RECIRCULATION PAGE 8 of 10 H_STEP ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED
- 11. CHECK VCT LEVEL GREATER THAN 12%
- Do the following:
EJ a) Increase VCT level to greater than 12%. b) Close Charging Pump Suction From RWST Isolation Valves:
- 1-CH-MOV-1115B D
- 1-CH-MOV-1115D LJ IF VCT level cannot be increased to greater than 12%, THEN locally unlock and close 1 -SI-46, RWST to Suction of Charging Pumps and Hydro Test Pump (located in the Auxiliary Building Basement near Unit 1 Penetration Area entrance).
- 12. VERIFY 1-E-0, ATTACHMENT 3, PRIMARY D Initiate ATTACHMENT 2, PRIMARY PLANT PLANT VENTILATION ALIGNMENT - VENTILATION ALIGNMENT.
COMPLETED
- 13. INITIATE ATTACHMENT 1, RECIRCULATION SPRAY SUMP PH ADJUSTMENT
- 14. VERIFY CONTAINMENT SUMP BORON CONCENTRATION:
a) Have Station Emergency Manager direct Plant Staff to initiate required Containment Sump sampling for boron concentration D b) Containment Sump boron concentration - D b) Initiate emergency boration. GREATER THAN THAT REQUIRED BY 1-SC-3.2, MINIMUM SHUTDOWN BORON CONCENTRATION FOR HZP, IZP AND CZP VS. BURNUP
NUMBER PROCEDURE TITLE REVISION 22 1 -ES-i 3 TRANSFER TO COLD LEG RECIRCULATION PAGE 9 of 10 ACTION! EXPECTED RESPONSE H RESPONSE NOT OBTAINED
- 15. DETERMINE IF TRANSFER TO HOT-LEG RECIRCULATION WILL BE REQUIRED:
D a) Check entry condition for this procedure - a) Do the following: FROM i-E-1, LOSS OF REACTOR SECONDARY COOLANT 1) WHEN the following conditions are met, THEN GO TO 1-ES-i .4, TRANSFER TO HOT LEG
\i RECIRCULATION, STEP 1:
D RCS subcooling based on core exit TCs is less than 25°F [75°F] AND D
- 4 hours 30 mm. have elapsed since event initiation AND D
- SI flow is indicated AND D
- Concurrence from TSC or Plant Staff
- 2) Evaluate and perform long-term operations:
- Consult TSC or Plant Staff D
- Determine if redundant SI Pumps can be stopped (STEP 15 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 22 1 -ES-i 3 TRANSFER TO COLD LEG RECIRCULATION PAGE 10 of 10 -_STEP_[H ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED
- 15. DETERMINE IF TRANSFER TO HOT-LEG RECIRCULATION WILL BE REQUIRED:
(Continued) C 3) Before 5 hours have elapsed following establishing Hot leg recirculation, GOTO 1-ES-i .5, TRANSFER FROM HOT LEG RECIRCULATION TO COLD LEG RECIRCULATION, STEP 1, to align Cold leg reci rcu lation.
- 4) Evaluate and perlorm long-term operations:
C
- Consult TSC or Plant Staff C
- Determine if redundant SI Pumps can be stopped C 5) Before 5 hours have elapsed following establishing Cold leg recirculation, GO TO i-ES-i .4, TRANSFER TO HOT LEG RECIRCULATION, STEP 1, to align Hot leg recirculation.
C 6) Repeat Step i5.a.2 through Step i5.a.5 RNOs, while continuing with procedure and step in effect. C b) RETURN TO procedure and step in effect
- END -
CONTINUOUS ACTION PAGE FOR 1-ES-1.3 NOTE: This Continuous Action Page is NOT in effect until STEP 9 is complete. NOTE: This Continuous Action Page remains in effect during sump recirculation. A , NOTE: IF Sump Blockage OR loss of suction occurs, THEN FRs should NOT be implemented until directed in ATTACHMENT 3, CONTAINMENT SUMP STRAINER BLOCKAGE OR LOSS OF SUCTION OR 1 -ECA-1 .1, LOSS OF EMERGENCY COOLANT RECIRCULATION. e 2 WHEN COLD LEG RECIRCULATION IS ALIGNED AND STEP 9 IS COMPLETED, THEN MONITOR THE FOLLOWING FOR INDICATIONS OF SUMP BLOCKAGE OR LOSS OF SUCTION: D
- LOW-HEAD SI PUMPS:
- Amps Oscillating
- Flow Oscillating D
- CHARGING PUMPS:
- Amps Oscillating
- Flow Oscillating
- Discharge Pressure Oscillating D
- INSIDE RECIRC SPRAY PUMPS:
- Oscillating:
- Amps
- Flow
- Discharge pressure
- Cavitation Annunciator Panel J 8-5
- - LIT D
- OUTSIDE RECIRC SPRAY PUMPS:
- Oscillating:
- Amps
- Flow
- Discharge pressure
- Cavitation Annunciator Panel J B-6 LIT D
- CONTAINMENT PRESSURE - INCREASING, due to loss of Recirc Spray
- 2. H SUMP BLOCKAGE OR LOSS OF SUCTION OCCURS, THEN PERFORM 1-ES-i .3, TRANSFER TO COLD LEG RECIRCULATION, ATTACHMENT 3, CONTAINMENT SUMP STRAINER BLOCKAGE OR LOSS OF SUCTION.
J2A)
NUMBER ATTACHMENT TITLE ATTACHMENT 1-ES-1.3 3 CONTAINMENT SUMP STRAINER BLOCKAGE OR REVISION LOSS OF SUCTION PAGE 22 lof2 CAUTION: If the suction source is lost to any SI Pump or Recirc Spray Pump, then the pump should be stopped. NOTE: This attachment remains in effect during sump recirculation.
- 1. IF any Low-Head SI Pump indicates sump blockage OR loss of suction, THEN do the following:
a) Place all non-running Charging Pumps in PTL. b) Place running Charging Pumps in PTL. c) Stop one affected Low-Head SI Pump. d) Monitor running Low-Head SI Pump. e) IF sump blockage OR loss of suction is indicated on running Low-Head SI Pump, THEN do the following:
- 1) Stop running Low-Head SI Pump.
- 2) Ensure Charging Pump Suction From RWST Isolation Valves are closed:
1-CH-MOV-1115B
- 1-CH-MOV-1115D
- 3) GOTOStep2.
f) IF running Low-Head SI Pump indications are normal, THEN do the following:
- 1) Ensure Charging Pump Suction From RWST Isolation Valves are closed:
- 1-CH-MOV-1115B
- 1-CH-MOV-1115D
- 2) Evaluate starting one Charging Pump.
- 3) Implement FRs as applicable.
NUMBER ATTACHMENT TITLE ATTACHMENT 1-ES-1.3 3 CONTAINMENT SUMP STRAINER BLOCKAGE OR LOSS OF REVISION PAGE SUCTION 22 2of2
- 2. HZ BOTH Low-Head SI Pumps are NOT available, THEN do the following:
a) Evaluate initiation of the following contingency actions: D
- Charging Pump Cross-Connect using O-AP-48, CHARGING PUMP CROSS CONNECT Make up to RWST as required:
D
- Make up with Unit 1 Blender using 1-OP-7.7, REFUELING WATER STORAGE TANK SYSTEM OPERATION OR D
- Make up with Unit 2 Blender using 1-OP-7.7, REFUELING WATER STORAGE TANK SYSTEM OPERATION OR EJ
- Consult TSC or Plant Staff for guidance to make up from Unit 2 RWST by cross-connecting RWSTs using RP System b) Do the following:
- Ensure ATTACHMENT 2, PRIMARY PLANT VENTILATION ALIGNMENT is initiated
- GO TO 1-ECA-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION, STEP 1
-END 7 1t f\e/v
(
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 74. WE 1 5EK2.2 074/NEW//H14/2 .7/2.9/4/
Given the following conditions:
- Unit 1 was initially at 100% power.
- B SG faulted inside Containment.
- The crew is unable to terminate SI, and have determined that B SG is also ruptured.
The crew has completed the appropriate EOP transitions, and is now performing 1-ECA-3.1, SGTR With Loss of Reactor Coolant Subcooled Recovery Desired. The STA reports the following:
- RWST Level is 38% and slowly decreasing.
- There is an ORANGE Path on Containment Sump Level (sump level is 11 feet and slowly increasing).
- There are no other RED or ORANGE paths.
Which ONE of the following identifies the appropriate procedure transitions for these conditions? A. DO NOT transition to 1-FR-Z.2, Response to High Containment Sump Level; remain in 1-ECA-3.1 and continue recovery actions. B. DO NOT transition to 1-FR-Z.2, Response to High Containment Sump Level; transition to 1-ES-i .3, Transfer to Cold Leg Recirculation. C. Transition to 1-FR-Z.2, Response to High Containment Sump Level, to identify the source of water: then return to 1-ECA-3.1 and continue recovery actions. D. Transition to 1-FR-Z.2, Response to High Containment Sump Level, to identify the source of water; transition to 1-ES-i .3, Transfer to Cold Leg Recirculation.
- a. Incorrect. Plausible since some EOPs preclude/delay implementation of FRPs either for reasons of priority or because specific ones are an expected consequence of the event in progress, ECA-3. i however has no exclusions.
- b. Incorrect. First part plausible as described above. Second part is plausible if the candidate is unsure of requirements for ES-i .3 implementation.
- c. Correct. As discussed above there are no exclusions/provisions for FRP implementation with ECA-3.1 in effect so Z.2 transition is required. Based on the conditions given the operator will return to ECA-3. 1 since based on the given conditions neither SG inventory, nor long term recirc capability are challenged at this time.
- d. Incorrect. First part is correct as discussed in Distractor c. Second part is incorrect but plausible as discussed above.
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3 Containment Flooding Knowledge of the interrelations between the (Containment Flooding) and the following: (CFR: 41.7 / 45.7) Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility. Tier: 1 Group: 2 Technical
Reference:
EOP ECA-3.1 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
DOMINION OP-AP-104 REVISION 2 PAGE 13 OF 20
- e. Initiate a Condition Report (CR) in order to evaluate the effectiveness of transient response procedures.
3.5.19 Except at Millstone Unit 2, Adverse Containment Setpoints
- a. Adverse Containment setpoints are entered based on plant specific setpoints. For North Anna, Surry and Kewaunee High Containment Pressure and High Containment Radiation are utilized. For Millstone Unit 3 High Containment Temperature and High Containment Radiation are utilized.
- b. Adverse setpoints are designated by parenthesis or brackets next to the normal setpoint.
- c. When the initiating setpoint is reached for Containment High Pressure or Containment High Temperature and then decreases to less than the setpoint Adverse number are no longer required. When the setpoint is reached for Containment High Radiation and then decreases to less than the setpoint, Adverse numbers are still required until an evaluation is performed on the effects to the instrument.
3.6 Except at Millstone Unit 2, Critical Safety Function (CSF) Status Trees Monitoring of CSF Status Trees is a crew responsibility. In the absence of a STA, a licensed Operator is assigned to monitor the CSF Status Trees. Millstone Unit 2 Monitor Resource Assessment Trees (RATs) in accordance with:
- EOP 2540-002 Resource Assessment Trees
- EOP 2540 Safety Function Tracking Page Procedure User 3.6.1 MONITOR Critical Safety Status carefully.
- a. Monitoring and implementation of the CSF Status Trees shall begin when directed by the initial emergency response procedure, or when a transition is made to another emergency procedure unless otherwise directed.
- b. Upon power restoration following a loss of all AC event, once any equipment has been placed in a disabled condition recovery efforts should be in accordance with the Loss of All AC guidelines. This is necessary because actions taken after equipment has been placed in disabled must be carefully restored to prevent damage to equipment and to align the emergency equipment as required for further EOP implementation. FRs will not apply until procedurally specified.
INFORMATION USE
DOMINION OP-AP-104 REVISION 2 PAGE 14 OF 20
- c. The CSF Status Trees have different rules of usage than emergency operating procedures and are monitored, at least every 15 minutes, in parallel with the performance of the emergency operating procedures.
- d. CSF Status Tree monitoring shall be continuous (approximately every 3 to 5 minutes) when a RED or ORANGE terminus is encountered.
- e. The SRO in charge of the Control Room activities shall be immediately informed if a RED or ORANGE terminus exists. The SRO in charge of the Control Room activities shall also be regularly advised of YELLOW or GREEN plant conditions.
- f. The CSF Status Tree shall be entered at the left side of the tree and each question of the tree branch shall be answered based on the existing Unit conditions.
- g. Each CSF Status Tree shall be monitored to completion at the trees terminus.
- h. The appearance of a RED or ORANGE path CSF Status Tree usually implies that some Unit equipment is not available or is significantly degraded.
- i. If an E, ES, or ECA series emergency operating procedure is suspended to perform a RED or ORANGE path Functional Restoration Procedure (FR), Operator judgement is required in subsequent procedures to avoid inadvertent reinstatement of a RED or ORANGE path by undoing a critical step of the original FR.
- j. If during the performance of any RED-condition FRG, a RED-condition of higher priority arises, then the higher priority condition should be addressed first, and the lower priority RED-condition FRG suspended.
Otherwise the FRG must be performed to the point of the defined transition regardless of whether the RED-condition has been cleared. After the new FRG has been completed and the guidance is to return to the procedure and step in effect, the transition should be to the previous FRG which has been implemented. In general, the performance of the critical safety functions is dependent on the current plant parameters. If a RED or ORANGE priority condition comes in and clears, the FRG does not need to be performed. If conditions degrade, the status of the safety function will become a continuous RED or ORANGE condition; at which time, the operator would be directed to the appropriate CSF.
- k. If an ORANGE path is encountered, the remaining CSF Status Trees shall be monitored. If a RED path is not encountered, the Recovery Procedures in progress shall be suspended and the FR required by the ORANGE path shall be performed.
INFORMATION USE
DOMINION OP-AP-104 REVISION 2 PAGE 15 OF 20 I. If a RED path or higher priority ORANGE path arises during the performance of any ORANGE path FR, then the RED path or higher priority ORANGE path should be addressed first and the lower priority ORANGE path FR suspended unless specifically addressed by the original FR (e.g., FR-C.2, Caution before step 11).
- m. FRs entered from a RED or ORANGE path shall be performed to completion unless pre-empted by a higher priority path or a loss of all AC power.
- n. A YELLOW terminus does not require immediate Operator attention. It is frequently indicative of an off-normal or temporary condition which will be restored to normal by actions that are already in progress.
- 1. A YELLOW path may provide an early indication of a developing RED or ORANGE path condition.
- 2. For YELLOW path conditions only, the cognizant SRO may decide whether or not to implement any YELLOW path condition FRs. Unit conditions should be evaluated to determine if implementation of the FR is appropriate.
- o. After restoration of the CSF Status Trees from a RED or ORANGE path condition, recovery actions may continue when the FR is completed.
Usually the FR will return the Operator to the procedure and step in effect. The Operator may be directed to another procedure due to conditions created during performance of the FR. NOTE:The Recovery Procedures are optimal assuming that equipment is available as required for safety. 3.6.2 MAKE any required adjustments to the Recovery Procedures based on equipment failures or unavailability. 3.6.3 If loss of AC power to the emergency busses occurs and ECA performance begins, none of the FR5 can be implemented because none of the electrically powered equipment used to restore a Critical Safety Function will be operable. 3.6.4 Certain procedures (e.g., ESs and ECAs) take precedence over the FR5. Typically, a NOTE will notify the Operator not to implement the FR under specific conditions. A INFORMATION USE
CONTINUOUS ACTION PAGE FOR 1-ECA-3.i
- 1. ADVERSE CONTAINMENT CRITERIA IF either of the following conditions exist, THEN use setpoints in brackets:
D
- 20 psia Containment pressure, OR D
- Containment radiation has reached or exceeded 1 .0E5 R/hr (70% on High Range Recorder).
- 2. SI REINITIATION CRITERIA IF either condition listed below occurs, THEN manually start Charging Pumps and align BIT:
D
- RCS subcooling based on Core Exit TCs LESS THAN 25°F [75°F], OR D
- PRZR level CANNOT BE MAINTAINED GREATER THAN 21%[26%].
- 3. SECONDARY INTEGRITY CRITERIA IF either of the following conditions exist AND the affected SG has NOT been isolated, THEN GO TO 1 -E-2, FAULTED STEAM GENERATOR ISOLATION, STEP 1, unless needed for RCS cooldown:
- Any SG pressure is decreasing in an uncontrolled manner, OR D
- Any SG has completely depressurized.
- 4. COLD LEG RECIRCULATION TRANSFER CRITERIA D RWST level decreases to less than 23%, THEN GO TO 1-ES-i .3, TRANSFER TO COLD LEG RECIRCULATION, STEP i.
- 5. ECST LEVEL CRITERIA C WHEN the ECST level decreases to 40%, THEN initiate i-AP-22.5, LOSS OF EMERGENCY CONDENSATE STORAGE TANK 1 -CN-TK-i.
- 6. RCP CRITERIA C
- Seal injection flow should be maintained to all RCPs.
C
- RCPs should be run in the following order of priority to provide PRZR spray: C, A C
- WHEN an RCP is stopped, THEN isolate PRZR spray from the stopped RCP.
- 7. QS TERMINATION CRITERIA C WHEN RWST level is less than 3% AND QS Pump amps are FLUCTUATING, THEN perform ATTACHMENT 6, TERMINATION OF QUENCH SPRAY.
- 8. REACTIVITY CONTROL CRITERIA C An Operator should be sent to locally close and lock i-CH-217, PG to Blender Isolation Valve.
S S C
QUESTIONS REPORT for NAPS 2010 RO NRC Exam rev3
- 75. WE16EG2.4.50 075/BANK/NAPS/L/3/4.2/4.O//
Given the following conditions:
- Unit 1 core off-load is in progress.
- A recently-irradiated fuel assembly drops while placing it into the containment upender.
- Bubbles are coming to the surface of the reactor cavity.
The following radiation monitors are in alarm:
- 1-RM-RMS-159, Containment Gaseous R/M.
- 1-RM-RMS-160, Containment Particulate R/M.
- 1-RM-RMS-162, Manipulator Crane Area RIM.
Which ONE of the following describes actions that will initially be required in accordance with O-AP-30, Fuel Failure During Handling? A. Place Fuel Building ventilation in service through the charcoal filters; evacuate the containment. B. Place Fuel Building ventilation in service through the charcoal filters; evacuate the Fuel Building. C. Manually initiate Control Room bottled air dump; evacuate the containment. D. Manually initiate Control Room bottled air dump; evacuate the Fuel Building.
- a. Incorrect. Plausible since there is a release potential and the candidate may conclude using the filters is the appropriate course of action. Second part is correct.
- b. Incorrect. First part incorrect but plausible as noted above. Second part is incorrect but plausible again based on release potentials and since the event is in containment the novice may rationalize that the correct answer wouldnt be true because it implies no one would be left in CNTMT to mitigate the event.
- c. Correct. First part is correct and required by AP-30 based on the given conditions. Second part correct in accordance with AP-30 as described above.
- d. Incorrect. First part is correct as discussed in Distractor c. Second part is incorrect but plausible as discussed in distractor b.
QUESTIONS REPORT for NAPS 2010 RD NRC Exam rev3 High Containment Radiation Ability to verify system alarm setpoints and operate controls identified in the alarm response manual. (CFR: 41.10 /43.5 /45.3) Tier: Group: Technical
Reference:
0-AP-30 Proposed references to be provided to applicants during examination: None Learning Objective: Question History: additional info:
U U
Doiiot NORTH ANNA POWER STATION ABNORMAL PROCEDURE NUMBER PROCEDURE TITLE REVISION 13 O-AP-30 FUEL FAILURE DURING HANDLING (WITH TWO ATTACHMENTS) PAGE 1 of 13 PURPOSE To provide instructions to follow in the event of a fuel failure during handling. ENTRY CONDITIONS This procedure is entered when any of the following conditions exist:
- Gas bubbles or discoloration of the water in the area of a fuel assembly, or
- Report of a fuel failure from refueling personnel, or
- Increasing radiation level on any of the following:
- 1-RM-RMS-159, Unit 1 Containment Gaseous
- 1-RM-RMS-160, Unit 1 Containment Particulate
- 1 -RM-RMS-1 62, Unit 1 Manipulator Crane
- 2-RM-RMS-259, Unit 2 Containment Gaseous
- 2-RM-RMS-260, Unit 2 Containment Particulate
- 2-RM-RMS-262, Unit 2 Manipulator Crane
- 1-RM-RMS-152, New Fuel Storage Area
- 1-RM-RMS-153, Spent Fuel Pit Bridge Crane CONTINUOUS USE
NUMBER PROCEDURE TITLE REVISION 13 O-AP-30 FUEL FAILURE DURING HANDLING PAGE 2 of 13 STEP ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED 1._ NOTIFY CONTROL ROOM NOTE: The Bases for Tech Spec 3.7.10 and 3.7.11 define recently irradiated fuel as fuel that has occupied part of a critical reactor core within the previous 300 hours.
- 2. CHECK IF MANUAL CONTROL ROOM BOTTLE AIR DUMP AND ISOLATION IS REQUIRED:
D a) Movement of recently irradiated fuel - D a) IF SRO desires Control Room bottle air IN PROGRESS dump and isolation, THEN continue with Step 2c. b) Either of the following - SATISFIED: 11 b)GOTOStep3. D
- Verbal report of a fuel handling accident OR D
- Any containment radiation monitor -
ii. IN ALARM (STEP 2 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 13 O-AP-30 FUEL FAILURE DURING HANDLING PAGE 3 of 13 ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED
- 2. CHECK IF MANUAL CONTROL ROOM BOTTLE AIR DUMP AND ISOLATION IS REQUIRED: (Continued) c) Place the following UNIT 1 CONTROL RM BOTTLED AIR SYS PANEL control switches in OPEN:
PANEL 1A, 1-EI-CB-156A D
- 1-HV-SOV-1300A 1-HV-SOV-1308A V D
- 2-HV-SOV-2300A PANEL 1B, 1-EI-CB-156B D
- 1-HV-SOV-1300B 4V
- 2-HV-SOV-2300B 2-HV-SOV-2308A d) Verify Control Room Damper status: C d) Close dampers.
C
- 1-HV-AOD-161-1 - CLOSED C
- 1-HV-AOD-160-1 -CLOSED C
- 1-HV-AOD-161-2 - CLOSED C
- 1-HV-AOD-160-2 - CLOSED (STEP 2 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 13 O-AP-30 FUEL FAILURE DURING HANDLING PAGE 4 of 13 ACTION! EXPECTED RESPONSE RESPONSE NOT OBTAINED H_STEP
- 2. CHECK IF MANUAL CONTROL ROOM BOTTLE AIR DUMP AND ISOLATION IS REQUIRED: (Continued) e) Place the following UNIT 2 CONTROL RM BOTTLED AIR SYS PANEL control switches in OPEN:
PANEL 2A, 2-EI-CB-156A D
- 1-HV-SOV-1300C 1-HV-SOV-1308B EJ
- 2-HV-SOV-2300C PANEL 2B, 2-EI-CB-156B D
- 1-HV-SOV-1300D D
- 2-HV-SOV-2300D 2-HV-SOV-2308B f) Verify Control Room Fan Status:
D 1) 1-HV-F-15 - OFF D 1) Stop 1-HV-F-15 1 2) 1 -HV-F-41 - ON D 2) Start 1 -HV-F-41 D 3) 2-HV-F-41 - ON D 3) Start 2-HV-F-41 g) Verify fan status on Auxiliary Shutdown D g) Start fans. Panels: C
- 1-HV-F-42-ON C
- 2-HV-F-42 - ON NOTE: If electrical power is unavailable to the Manipulator Crane, then manual operation may be required.
- 3. PLACE THE FAILED FUEL ASSEMBLY IN A SAFE LOCATION
- 4. EVACUATE AFFECTED AREA
(,VV.
NUMBER PROCEDURE TITLE REVISION 13 O-AP-30 FUEL FAILURE DURING HANDLING PAGE 5 of 13 j_STEP ACTION! EXPECTED RESPONSE I RESPONSE NOT OBTAINED NOTE: For a Fuel Handling Accident in Containment, the required closure actions will be based on the existing conditions and established radiation protection practices. When acceptable radiological protection conditions exist for closure team personnel, containment closure will be established within 45 minutes.
- 5. CHECK ACCIDENT - IN FUEL BUILDING Do the following:
D a) Initiate ATTACHMENT 2, CLOSING AIRLOCK DOORS IN AN EMERGENCY for affected Containment. b) Verify Containment Closure:
- 1) Refer to the applicable Containment Boundary Breach Log, for open penetrations:
D
- 1-LOG-18 D
- 2-LOG-18 D 2) Notify Health Physics.
Q 3) Establish containment closure for open penetrations.
- 6. NOTIFY HEALTH PHYSICS DEPARTMENT
- 7. CHECK IF ACCIDENT IS IN UNIT 1 GOTO Step 10.
CONTAINMENT
z z LU 0 LU Co H CDco 0 >.,- 4: C LU 0 (0 a: C-) ci) U) 2 cci C\1 o cJ, C 0 (ciH ci) U) 0 z
-J P
0 LU 0 LU z 0 -J 4: I D 0 H H 0 LU z a: a: D D z 0 0 LU LU 4: C) a: 0 D oz a: -J 0 4: U-
-J LU LUO D
LL HO z a: a: 0 0 0
- 0) CJ U, (0 (0 r
4: H th th th Da: a: a: a: QH LUZU) a: 0 r05 1 LU ) 1 C 0W U 0 0 0 D z 0 C
NUMBER PROCEDURE TITLE REVISION 13 O-AP-3O FUEL FAILURE DURING HANDLING PAGE 7 of 13 j_STEP_F] ACTION/EXPECTEDRESPONSE I RESPONSENOTOBTAINED I NOTE: A HI-RAD alarm on 1-RM-RMS-159, 1-RM-RMS-160, or 1-RM-RMS-162 will trip Containment Purge Supply and Exhaust Fans to Unit 1 Containment unless Unit 2 Containment is being ventilated.
- 9. VERIFY AUTOMATIC ACTUATIONS HAVE OCCURRED ON CONTAINMENT RADIATION MONITOR ALARM:
a) Containment Purge Fans TRIPPED:
- C a) IF Unit 2 Containment is being ventilated, THEN continue to operate Containment C
- 1-HV-F-4A Supply Purge Fans.
C
- 1-HV-F-4B Supply C
- 1-HV-F-5A Exhaust C
- 1-HV-F-5B Exhaust b) Isolation dampers on Purge Supply and C b) Close any open damper.
Exhaust CLOSED: C
- 1-HV-MOV-100A C
- 1-HV-MOV-100B C
- 1-HV-MOV-100C C
- 1-HV-MOV-100D C
- 1-HV-MOV-1O1 C
- 1-HV-MOV-102 C c) GOTO Step 16 1O. CHECK IFACCIDENT IS IN UNIT2 C GOTOStep13.
CONTAINMENT
NUMBER PROCEDURE TITLE REVISION 13 O-AP-30 FUEL FAILURE DURING HANDLING PAGE 8 of 13 j_STEP_fI ACTION/EXPECTED RESPONSE I RESPONSE NOTOBTAINED I
- 11. CHECK UNIT 2 CONTAINMENT RADIATION D HZ a high radiation alarm is received, THEN MONITORS ANY HIGH RADIATION ALARM RETURN TO Step 2.
EXISTS: D
- 2-RM-RMS-259 D GO TO Step 16.
OR D
- 2-RM-RMS-260 OR D
- 2-RM-RMS-262 NOTE: A HI-RAD alarm on 2-RM-RMS-259, 2-RM-RMS-260, or 2-RM-RMS-262 will trip Containment Purge Supply and Exhaust Fans to Unit 2 Containment unless Unit 1 Containment is being ventilated.
- 12. VERIFY AUTOMATIC ACTUATIONS HAVE OCCURRED ON CONTAINMENT RADIATION MONITOR ALARM:
a) Verify Containment Purge Fans - lJ a) IF Unit 1 Containment is being ventilated, TRIPPED: THEN continue to operate Containment Purge Fans. D
- 1 -HV-F-4A Supply D
- 1-HV-F-4B Supply D
- 1-HV-F-5A Exhaust D
- 1-HV-F-5B Exhaust (STEP 12 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 13 O-AP-30 FUEL FAILURE DURING HANDLING PAGE 9 of 13 H_STEP ACTION/EXPECTEDRESPONSE I RESPONSENOTOBTAINED I
- 12. VERIFY AUTOMATIC ACTUATIONS HAVE OCCURRED ON CONTAINMENT RADIATION MONITOR ALARM: (Continued) b) Verify Isolation Dampers on Purge Supply EJ b) Close any open damper.
and Exhaust CLOSED:
- 2-HV-MOV-200A D
- 2-HV-MOV-200B D
- 2-HV-MOV-200C D
- 2-HV-MOV-200D D
- 2-HV-MOV-201 D
- 2-HV-MOV-202 D c) GOTO Step 16
- 13. CHECK FUEL BUILDING RADIATION U WHEN a high radiation alarm is received, MONITORS ANY HIGH RADIATION ALARM THEN perform Step 14.
EXISTS: U
- 1-RM-RMS-152,NEWFUEL U GOTOStep16.
STORAGE OR U
- 1-RM-RMS-153, SPENT FUEL PIT BRIDGE CRANE
NUMBER PROCEDURE TITLE REVISION 13 O-AP-30 FUEL FAILURE DURING HANDLING PAGE 10 of 13 J_STEP ACTION! EXPECTED RESPONSE I RESPONSE NOT OBTAINED CAUTION: To prevent entry of contaminants into the Control Room, the Hi-Hi alarm on 1-RM-RMS-152 and 1-RM-RMS-153 should not be reset until Health Physics has verified that the Turbine Building atmosphere is acceptable. 14._ VERIFY CONTROL ROOM ISOLATION - ACTUATED: a) Fuel Building Radiation Automatic Interlock D a) IF SRO desires Control Room bottle air key switch in ENABLE
- dump and isolation, THEN continue with Step 14b.
NOT, THEN GOTO Step 15. b) Verify automatic dump of Bottled Air E1 b) Manually dump Control Room Control System ON:
- Room Bottled Air System.
D
- 1-HV-Pl-1311 D
- 2-HV-PI-231 1 c) Verify Control Room Fan Status:
D 1) 1-HV-F-15 - OFF D 1) Stop 1-HV-F-15 D 2) 1-HV-F-41 - ON 2) Start 1-HV-F-41 D 3) 2-HV-F-41 - ON EJ 3) Start 2-HV-F-41 (STEP 14 CONTINUED ON NEXT PAGE)
NUMBER PROCEDURE TITLE REVISION 13 O-AP-30 FUEL FAILURE DURING HANDLING PAGE 11 of 13 ACTN/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 14. VERIFY CONTROL ROOM ISOLATION -
ACTUATED: (Continued) d) Verify Control Room Damper status: 1J d) Close dampers. D
- 1-HV-AOD-161-1 -CLOSED D
- 1-HV-AOD-160-1 - CLOSED D
- 1-HV-AOD-161-2 - CLOSED D
- 1-HV-AOD-160-2 - CLOSED e) Verify fan status on Auxiliary Shutdown D e) Start Fans.
Panels:
- 1-HV-F-42 - ON D
- 2-HV-F-42 ON -
- 15. _VEIFY FUEL BUILDING VENTILATION D Align Fuel Building ventilation through the
( ALIGNED: __-D a) Fuel-Btilding ExhatFan AT LEAST D charcoal filter GOTOStep16. ONEON: -
//
U
- 1 -HV-F-7A Fuel BuiMing Supply Fan 1-HF-OFF (STEP 15 CON
NUMBER PROCEDURE TITLE REVISION 13 O-AP-30 FUEL FAILURE DURING HANDLING PAGE 12 of 13 -j_STEP fj ACTION/EXPECTEDRESPONSE I I RESPONSENOTOBTAINED
- 15. VERIFY FUEL BUILDING VENTILATION -
ALIGNED: (Continued) c) Check either set of Auxiliary Building Iodine Filter Dampers in FILTER: D
- 1-HV-AOD-107A1,2,3,4 OR D
- 1-HV-AOD-107B 1,2,3,4 d) Fuel Building Exhaust aligned to charcoal filter 1-HV-FL-3A or 1-HV-FL-3B as indicated by the following dampers in F I LTER:
L1
- 1-HV-AOD-107-1 D
- 1-HV-AOD-107-2 D
- 1-HV-AOD-107-3 D
- 1-HV-AOD-107-4
- 16. INITIATE EPIP-1.O1, EMERGENCY MANAGER CONTROLLING PROCEDURE
NUMBER PROCEDURE TITLE REVISION 13 O-AP-30 FUEL FAILURE DURING HANDLING PAGE 13 of 13 -j_STEP1I ACTION! EXPECTED RESPONSE I RESPONSE NOT OBTAINED
- 17. ALIGN THE CONTROL ROOM EMERGENCY VENTILATION SYSTEM WITHIN ONE HOUR OF EVENT:
a) Place ONE of the following Emergency Ventilation Fans in service on Turbine Building supply using O-OP-21 .7, Main Control Room and Relay Room Emergency Ventilation Operation: D
- 2-HV-F-41 D
- 1-HV-F-42 C
- 2-HV-F-42 C b) Verify all other available Emergency Ventilation fans are in service on recircu lation
- 18. CHECK IF CAUSE OF ACCIDENT - C WHEN cause of accident has been CORRECTED corrected, THEN continue with Step 19.
- 19. RESTORE VENTILATION SYSTEMS TO NORMAL USING:
C
- O-OP-21.7, MAIN CONTROL ROOM AND RELAY ROOM EMERGENCY VENTILATION OPERATION C
- 1-OP-21.2,CONTAINMENTPURGE C
- 2-OP-21.2, CONTAINMENT PURGE
- 20. RETURN TO PROCEDURE IN EFFECT
- END -
NUMBER ATTACHMENT TITLE ATTACHMENT 0-AP-30 1 REFERENCES REVISION PAGE 13 lof2
- 1. UFSAR Section 9.4.9
- 2. UFSAR Section 11.4.2.18
- 3. UFSAR Section 15.4.5
- 4. 11715-ESK-6KE
- 5. 11715-ESK-6KF
- 6. TRM 3.3.7
- 7. Tech Spec 3.7.10
- 8. TRM 3.9.5
- 9. Tech Spec 3.7.15
- 10. DCP 89-32
- 11. 1(2)-OP-21.2, CONTAINMENT PURGE
- 12. 0-OP-21 .7, MAIN CONTROL ROOM AND RELAY ROOM EMERGENCY VENTILATION OPERATION
- 13. EPIP-1.01, EMERGENCY MANAGER CONTROLLING PROCEDURE
- 14. Tech Spec Amendment 198, Containment Building Penetrations Unit 1
- 15. Tech Spec Amendment 179, Containment Building Penetrations Unit 2
- 16. DR N-99-0215, Control Room Pressurization, Isolation, and Ventilation
- 17. DCP 00-1 69, Control Room Bottled Air System Modification
- 18. TSCR N-01 1, Alternate Source Team (Rev. 11, Step 2.a and note, 14.a and Attachment 2, Step 1)
NUMBER ATTACHMENT TITLE ATTACHMENT 0-AP-30 1 REFERENCES REVISION PAGE 13 2of2
- 19. Tech Spec change N-052A, Deletion of MCR/ESGR Bottled Air System from Tech Specs (Rev. 12, Step 17)
- 20. TRM 3.3.9 Regulatory Guide (RG) 1 .97 Instrumentation, Tech Spec Bases 3.3.3 Post Accident Monitoring (PAM) Instrumentation, LA001345 NAPS Implementation Plan for TRCR 012, TSCR 023, & UFSAR FN 2008-008; Reg Guide 1.97, PAM Instrumentation (Rev. 13)
- 21. CAl 53970, CR359803, Unable to Perform Validation of TCA (Rev. 13, Step 2 and Step 1 4.d&e)
NUMBER ATTACHMENT TITLE ATTACHMENT O-AP-30 2 CLOSING AIRLOCK DOORS IN AN EMERGENCY REVISION PAGE 13 lofi Have the Maintenance Department install the Equipment Door and Temporary Penetration Plate using O-MCM-1204-05, EMERGENCY INSTALLATION OF EQUIPMENT DOOR AND TEMPORARY PENETRATION PLATE, as directed by Health Physics based on radiological conditions. CAUTION: At least one Personnel Airlock door must be closed. The inner door is the preferred door. If possible, then both doors should be closed.
- 2. Ensure any covers placed over the door seating surfaces are removed.
a) Close the Personnel Airlock Door as follows: b) Verify door seating surfaces are clean. c) Coordinate with Control Room to manipulate ventilation as necessary to equalize pressure between the Auxiliary Building and Containment:
- Containment Ventilation Aux Building Supply Fan(s)
- Aux Building Exhaust Fan(s) d) Pull Inner Door closed and HOLD Inner Door in the CLOSED position until Locking Ring is engaged in the next step.
e) Press and hold the Inner Door CLOSE pushbutton until the Inner Door is fully closed. f) Press the Outer Door OPEN pushbutton to fully open the Outer Door Locking Ring. g) WHEN the Outer Door Locking Ring is full open, THEN close Outer Door and HOLD Outer Door in the CLOSED position until the Locking Ring is engaged in the next step. h) Press and HOLD the Outer Door CLOSE pushbutton until the Outer Door is fully closed. i) Notify Control Room of the status of Personnel Airlock doors.
- END -}}