Text

Farley Initial Exam 2010-301 Draft RO Written Exam (Section 2)
	ML101650244
Person / Time
Site:	Farley
Issue date:	06/07/2010
From:	; NRC/RGN-II
To:	; Southern Nuclear Operating Co
References
	50-348/10-301, 50-364/10-301
	Download: ML101650244 (558)
	v • d • e

{{#Wiki_filter:1. 001AK2.06 001/FNP BANK/RO/C/A 2.9/3.2IY 2007/N/2ICVRIY Unit 1 is at 74% power and stable, and the following conditions occurred: At 1000:

Rod control is in AUTO.
TI-408A, Tavg -Tref deviation, indicates O°F and stable.
Pressurizer level is stable.
Reactor Power is approximately 75% and stable.
Control Bank 0 step counters are at 144 steps. At 1002:
TI-408A, Tavg -Tref deviation, indicates approximately

+2°F and rising.

Pressurizer level is slowly rising.
Pressurizer spray valves have throttled open.
Reactor Power is approximately 76% and slowly rising.
Control Bank 0 step counters are at 150 steps and rising at 8 steps per minute.
There is no load change in progress.

Which one of the following describes:

1) the event in progress and 2) the NEXT action that must be performed lAW AOP-19.0, Malfunction of Rod Control System? A. 1) Inadvertent RCS boration;

2) Trip the reactor and enter EEP-O, Reactor Trip or Safety Injection.

B. 1) Inadvertent RCS boration;

2) Place the rod control mode selector switch to MANUAL and match Tavg with Tref by inserting rods. C. 1) Uncontrolled Continuous Rod Withdrawal;

2) Trip the reactor and enter EEP-O, Reactor Trip or Safety Injection. 1) Uncontrolled Continuous Rod Withdrawal;

2) Place the rod control mode selector switch to MANUAL and verify that rod motion stops. Page: 1 c:I 200 1211412009 A -Incorrect.

The first part is incorrect, since for an inadvertent boration, TavglTref mismatch would be less than -1.5 (with rods to be moving outward) and power would be less than 75% instead of 76%. Plausible, since rods would be moving out and TavglTref mismatch could be increasing (which would cause Przr level to rise and spray valves to throttle open) with an inadvertent boration. The second part is incorrect, but plausible. The stated action is the RNO if rods do not cease moving once they have been placed in manual lAW AOP-19. Also, a conservative action may be chosen to trip the reactor, but this would not be in accordance with AOP-19.0 for this situation, nor would it be necessary. B -The first part is incorrect (see A). Second part is correct lAW AOP-19 for a continuous rod withdrawal (see D). C -Incorrect. The first part is correct (see D). The second part is incorrect (see A). o -Correct. A CRW is taking place as indicated by the Tavg/Tref meter value going up above +1.5 and continuing to increase. This shows rods should actually be moving to lower the high temperature, and the action is to place rods in Manual if they are stepping while in AUTO. Technical

Reference:

AOP-19 Malfunction of Rod Control, Version 26.0 Previous NRC exam history if any: FNP 2007 NRC exam, but with different distractors (changed from inadvertent dilution to inadvertent boration in A & B). This is the only question in the bank that comes close to meeting this k/a (searched BOTH "KA" and "second KA" on "contains 001AK"). 001AK2.06 001 Conti nuous Rod Withdrawal AK2. Kno.vledge of the interrelations between the Continuous Rod Withdrawal and the follo.ving: (CFR 41.7 /45.7) AK2.06 T-ave./rEf. deviation ma:er ......................................... 3.0* 3.1 Match justification: This question presents conditions indicating a Continuous Rod Withdrawal, and the Tavg/Tref meter value and trend is provided. To obtain the correct answer, a knowledge of the relationship between the CRW and the Tavg/Tref meter response is required. Objective:

4. EVALUATE plant conditions and DETERM I NE if any system components need to be operated while performing AOP-19, Malfunction of Rod Control System. ( OPS-52520SOO)

Page: 2 of 200 1211412009 ...' i 10/27/09 10:06:24 FNP-1-AOP-19.0 MALFUNCTION OF ROD CONTROL SYSTEM Version 26.0 I I [ Action/Expected Response Response Not Obtained NOTE: Steps 1 and 2 are IMMEDIATE OPERA TOR actions. -1 Verify NO load change in progress. 1 Check for cause of load change. [YES[ 1.1 IF load rejection in progress or has occurred, THEN go to FNP-1-AOP-17.0, RAPID LOAD REDUCTION. 1.2 IF secondary leakage is indicated, THEN go to FNP-1-AOP-14.0, SECONDARY SYSTEM LEAKAGE. -2 IF unexplained rod motion occurring, THEN stop rod motion. CORRRECT 2.1 IF rod control in AUTO, 2.1 IF rod control in MANUAL, THEN place rod control in MANUAL. THEN place rod control in AUTO NOTE: In AUTO rod control, rods will step OUT ifTAVG less than TREF by at least 1.5 degrees, and Rods will step IN if T A VG greater than TREF by at least 1.5 degrees. 2.2 2.2.1 2.2.2 IF unexplained rod motion NOT stopped, THEN perform the following. IA&C I Trip the reactor IINCORRRECT Go to FNP-1-EEP-0, REACTOR TRIP OR SAFETY INJECTION Page Completed Page 2 of9 2.1.1 IF AUTO rod motion due to T A VG/TREF mismatch, THEN verify rod motion stops when TAVG is within 1 degree ofTREF /

2. 001K2.02 00l1MOJ:;!JRO/MEM 3.6/3.71IN/2ICVRNER 5 EDITORIAL Which one-of the following correctly describes components in the power flow path to the Reactor Trip Breakers?

The 600V (1) supply the CRDM MG set supply breakers, then the (2) ,then the Reactor trip breakers. (1 ) (2) A. LCCs 0 and E Motor Generator Sets, then the Power Cabinets B. MCCs A and B Motor Generator Sets, then the Power Cabinets LCCs 0 and E Motor Generator Sets D. MCCs A and B Motor Generator Sets A. Incorrect. Plausible because all parts of the correct answer are listed, but in an incorrect order. The power cabinet is actually downstream of the Reactor trip breaker. B. Incorrect. MCC A & B are incorrect, but Plausible; these are also safety related 600V switchgear. See A for second part. C. Correct. Per load list and FSD on Reactor Protection, A181007, Figure F-1. D. Incorrect. MCC A & B are incorrect, but Plausible; these are also safety related 600V switchgear. Each MG set consists of a600V AC, 150 hp induction motor, a stainless steel flywheel, and a 260V AC, 3 phase synchronous generator located in the non-rad Aux bldg 139' elev. The motors for the M G sets are powered from two 600V load centers (LCs) (M G A is powered from LCD; M G B is powered from L C E). These motor supply breakers can be operated from the M G set control panels located in the rod control room (Aux bldg 121' elev.) or from the MCB. Page: 3 of 200 1211412009 Previous NRC exam history if any: 2005 Vogtle NRC exam under 001 K2.01 (power supplies to MG SETS) 001 K2.02 001 Control Rod Drive System K2 KnoNledgeof buspo.ver supplies to the foilONi ng: (CFR: 41.7) K2.02 One-linediaaran of power supply to trip brEEKers ..................... 3.63.7 Match justification: The power flow to the Reactor trip breakers is examined here including the Busses that supply the CRDM MG. This power supply flow must be understood to correctly answer this question. Objective:

1. NAM E AND I DENTI FY the power supply for the following cabinets associated with the Reactor Protection System (RPS) to include those items found on Figure 12, 120 VAC Distribution (OPS-52201104).

Page: 4 of 200 1211412009 CONTROL { BOARD MASTER AND ACTUATE TRAIN 'B' SWITCHES OUTPUT SLAVE RELAYS SAFEGUARDS <TRAIN 'B') PROTECTION ANALOG PROTECTION SYSTEM SYSTEM LOGIC SOLID STATE LOGIC <'tI. 7300 SYSTEM) TRAIN B NUCLEAR INSTRUMENTATION SYSTEM / " INPUT DR FlELD;oNTACTS { PROCESS CHANNEL L __ -+ __ ____ __ L-__ 4-__ __ __ SENSOR COMPUTER IV*}----/ Il) BISTABLE I,.-'OR' CABLE CONTROL BOARD ____ CONTROL PROTECTION I I I BOARD SYSTEM I SOLID STATE LOGIC I DEMUX TRAIN A ___ -1 CABINET MASTER AND CONTROL SLAVE RELAYS 1"-ISOLA nON BOARD SWITCHES Urf---------t-- <TRAIN 'A') ACTUATE TRAIN 'A' SAFEGUARDS c ROD CONTROL SYSTEM I;:' !r ,TRIP 1 1\-( BRKR. r ( BYPASS "B' I BRKR. 'B' V REACTOR TRIP C"" J BRKR. (BYPASS , 'A' J BRKR. 'A' II CONTROL CONTROL M-G M-G SET A SETB> ) &£'0 vI.-L--I-,It{J <....--REACTOR PROTECTION SYSTEM BOUNDARIES fiGURE f-l FNPUNIT 1

LOAD LIST A-S062S0 DF03 10 600V LOAD AB -139' D177010 TPNS ED09 EDiO EDII NIT47MOOOIB-A ED15 QIE12MOOOIA-A ED16 QIE12MOOOIB-A lsectf.doc DESCRIPTION FAN SEE PAGE F-3 F-79 0-46 I SWQ1Rl >>> iC BAT.T F-3 CHARGER* (ATRAINStJPPLY)

>>> i2SVDC DISC SWQiR18B0002A-A >>>.lA12SVDC SWGR >>> lA 600 IB CRDM COOLER FAN iF 600V LOAD CENTER (ALTERNATE) <<< EF06 IA CONTAINMENT COOLER (EMERG. LOW SPEED) IB CONTAINMENT COOLER (EMERG./ LOW SPEED) F-92 F-98 F-I04 Page F -2 Ver. 47.0 FNPUNIT 1 LOAD LIST A-S062S0 --'"" DG03 IE 600V LOAD ') '---------- AB -121' D177011 BKR TPNS Q I R I6BKREEO I Q1RllB0005-B NICIIMOOOIB-N Q1R16B0005 .. B EEOS QIEI2MOOOIC-B EE09 QIE22MOOOIB-B EE10 EEll EE12 Q1R16B0008-AB EE13 NI T47MOOOIA-B EE14 EEl5 Q1R17B0009-B EEI6 QIEI2MOOOID-B lsectg.doc DESCRIPTION SEE PAGE G-3 AC SW >>>lCBATT G-3 CHARGER Q1R42E0001C.AB(B TRAIN*SUPPLY) >>> 125V DCDISe SWQ1R18B0002B':B >>> IB125V DCSWGR >>> le 600V LOAD (ALTERNATE-EMERG>>>> E-5 EC10>>> IC CONTAINMENT COOLER (EMERG./ LOW SPEED) IB REACTOR CAVITY DILUTION FAN G-76 MCC>>> G-S4 lF 600V LOAD CENTER (ALTERNATE) <<< EF08 IA CRDM COOLER FAN 1 T600/208V Mee >>> G-99 1 V600/208V MCe >>> G-105 I ID CONTAINMENT COOLER (EMERG./ LOW SPEED) Page G -2 Rev. 23

1. ROD CONT-40204I04 004IHLT//M (LEVEL 1) SYS/001K2.011111 Page: 1 001 K2.02 Which ONE of the following correctly states the order of components through which power flows to the Control Rod Drive Mechanisms?

A. MG Supply Breakers, then 600v LCC D and E, then Motor Generator Sets, then Power Cabinets, then Reactor Trip Breakers. B. MCC A and B, Power Cabinets, then MG Supply Breakers, then Motor Generator Sets, then Reactor Trip Breakers. C':'" 600v LCC D and E, then MG Supply Breakers, then Motor Generator Sets, then Reactor Trip Breakers, then Power Cabinets. D. MCC A and B, Motor Generator Sets, then MG Supply Breakers, then Reactor Trip Breakers, then Power Cabinets. A. Incorrect. Plausible because all parts of the correct answer are listed, but in an incorrect order. B. Incorrect. Plausible because all parts of the correct answer are listed, but in an incorrect order. C. Correct. See Reference 1, Page 9. D. Incorrect. Plausible because all parts of the correct answer are listed, but in an incorrect order. This choice would be correct if "Motor Breakers" were replaced with "Generator Breakers". Each MG set consists of a 600V AC, 150 hp induction motor, a stainless steel flywheel, and a 260V AC, 3 phase synchronous generator located in the non-rad Aux bldg 139' elev. The motors for the MG sets are powered from two 600V load centers (LCs) (MG A is powered from LC D; MG B is powered from LC E). These motor supply breakers can be operated from the MG set control panels located in the rod control room (Aux bldg 121' elev.) or from the MCB. 2005 VNP nrc exam KIA 001 Control Rod Drive K2.01 Knowledge of bus power supplies to the following: One-line diagram of power supply to MIG sets. KIA MATCH ANALYSIS Question tests knowledge of the power supplies to the MIG Sets at the memory level. 10126/2009

3. 003A2.03 2.7/3.1/N/N/4/CVRIY Unit 1 is at 25% power and the following conditions occurred:

At 1000:

1 A RCP amps and motor winding temperature were observed to be rising while 1A RCS LOOP flow was decreasing.

At 1002:

EF1, 1A RCS LOOP FLOW LO OR 1A RCP BKR OPEN, is in alarm.
1A RCP Handswitch indicating Green and Amber lights are LIT, the Red light is NOT LIT.
AOP-4.0, Loss Of Reactor Coolant Flow, immediate actions have been completed.

RCS Temperatures are: 1A RCS LOOP Tavg is 53rF. 1 B RCS LOOP Tavg is 553°F. 1 C RCS LOOP Tavg is 553°F. Which one of the following correctly describes the CAUSE of these indications and the ACTION required lAW AOP-4.0? CAUSE ACTION A'I Seized motor bearing Trip the Reactor B. Sheared shaft Trip the Reactor C. Seized motor bearing Commence Normal Reactor Shutdown D. Sheared shaft Commence Normal Reactor Shutdown Page: 5 of 200 1211412009 This is on the RO level, since TS 3.4.2 requires Mode 3 in 30 minutes for Tavg below Minimum Temperature for Criticality. AOP-4.0 requires a reactor trip in this situation at step 3 after immediate action steps 1 & 2. This would be SRO were .. .. 4 A -Correct. As the motor bearing starts to seize, the amps go up and flow goes down until the breaker trips on overcurrent. Then, AOP-4.0 entry conditions are met ("This procedure is entered when forced RCS flow is lost in one or more loops and no reactor trip is required.") AOP-4.0 requires a reactor trip if any Tavg is < 541°F, and the stem gives 539°F for A Loop. B -Incorrect. The first part is incorrect, but plausible. RCS flow would go down if the shaft sheared, but current would also go down instead of up. The second part is correct (see A). C -Incorrect. The first part is correct (see A). The second part is incorrect, but plausible. It would be correct at this power level <<30%, P-8) if Tavg was not less than the Minimum Temperature for Criticality. D -Incorrect. The first part is incorrect (see B). The second part is incorrect (see C). AOP-4, Version 18.0 TS 3.4.2 Page: 6 of 200 1211412009 Previous NRC exam history if any: 003A2.03 003 Reactor Cool ant Pump System A2 Ability to (a) predict the impacts of thefoilONing malfunctions or operations on the RepS; and (b) based on those predictions, use procedures to correct, control, or mitigate the conSlquences of those malfunctions or operations: (CFR: 41.5/43.5/45.3/45/13) A2.03 Problemsassxicted with RCP motors, including fruity motors and current, a1d winding and booring tempercture probl ems ....... 2.7 3.1 Match justification: This question requires knowledge of what type of motor malfunction would give the indications in the stem. The indications were given in the stem and the applicant is required to analyze and diagnose what malfunction would cause these indications to avoid backwards logic. This order was also required to allow choosing actions which are based on the indications. At FNP, Motor bearing temperature indication is not available, but motor winding temperatures are. Winding temperatures would go up due to the motor shaft and bearing seizing, so is included in the stem. The second part of the question and choices require knowledge of what action is required for the given set of indications. Objective:

6. DEFI NE AND EVALUATE the operational implications of normal / abnormal plant or equipment conditions associated with the safe operation of the Reactor Coolant Pumps (RCPS) components and equi pment, to i ncl ude the foil owi ng (OPS-40301 007):

Normal Control Methods
Abnormal and Emergency Control Methods
Protecti ve i sol ati ons such as hi gh flow, low pressure, low I eve! i ncl udi ng setpoi nt
Fast dead bus transfer
A utomati c actuati on i ncl udi ng setpoi nts
A cti ons needed to mi ti gate the consequence of the abnormal i ty Page: 7 of 200 1211412009 05/12/09 12:22:58 FNP-I-AOP-4.0 LOSS OF REACTOR COOLANT FLOW Action/Expected Response Response Not Obtained NOTE: Step 1 and 2 are IMMEDIATE OPERATOR actions. 1 Maintain SG narrow range level stable at approximately 65% using: [] Main Feedwater Regulating Valves [] Main Feedwater Bypass Regulating Valves. [] Auxiliary Feedwater Control Valves. 2 Check lA and IB RCPs -RUNNING. Monitor Tavg for all three RCS loops 2:: 541°F. (TS 3.4.2) (J 2 3 IF SG level rise cannot be controlled, THEN close the affected SG Main Feedwater Stop Valve(s) OR Auxiliary Feedwater Stop valve(s).

[ ] IA SG QIN21MOV3232A [ ] IB SG QIN21MOV3232B [ ] lC SG QIN21MOV3232C OR [ ] lA SG QIN23MOV3350A [ ] 1B SG QIN23MOV3350B [ ] lC SG QIN23MOV3350C Manually close pressurizer spray valve for affected RCP. [ ] lA RCS loop spray valve PK-444C [ ] 1B RCS loop spray valve PK-444D Perform the following .. IF the main generator is ON THEN trip the reactor and go to FNP-I-EEP-O, REACTOR SAFETY INJECTION 3.2 IF the main generator is OFF LINE, THEN raise Tavg 2:: 541°F within 30 minutes 3.2.1 Adjust steam dumps to reduce secondary power demand as necessary 3.2.2 Verify rod control in MANUAL o Step 3 continued on next page Page 2 of 11 05112/09 12:22:58 FNP-I-AOP-4.0 ',) JLJL LOSS OF REACTOR COOLANT FLOW Version 16.0 I I I Action/Expected Response Response Not Obtained 4 Maintain PRZR pressure 2200-2300 psig. 4.1 Control PRZR heaters as required. 4.2 IF lA and IB RCPs running, THEN, control pressurizer pressure with both normal spray valves. 3.2.3 Stabilize Tavg in the idle loop(s) > 541°F while maintaining the running loop(s) < 554°F by adjusting rod position and/or boron concentration 3.2.4 IF unable to restore Tavg 541°F, THEN trip the reactor and go to FNP-1-EEP-0, REACTOR TRIP OR SAFETY INJECTION 4.2 Perform the following 4.2.1 IF 1B RCP running, THEN control pressurizer pressure with PK-444D. NOTE: Running lA and lC RCPs will be required to provide adequate spray flow through the 1A RCS loop spray valve. 4.3 Proceed to step 5. Page Completed 4.2.2 IF 1A & lC RCPs are running, THEN control pressurizer pressure with PK-444C. 4.2.3 IF spray flow is adequate, THEN proceed to step 5. 4.2.4 IF no spray valves are available, THEN proceed to step 4.4. o Step 4 continued on next page Page 3 of 11 05112/0912:22:58 FNP-1-AOP-4.0 LOSS OF REACTOR COOLANT FLOW Version 16.0 Action/Expected Response 6 Maintain PRZR level at approximately 7 Within six hours of the loss of ReS complete the following: 7.1 IF the unit is in Mode 1 or 2, THEN place unit in Mode 3 using the following procedures: [] FNP-1-UOP-3.1, POWER OPERATION [] FNP-1-UOP-2.l, SHUTDOWN OF UNIT FROM MINIMUM LOAD TO HOT STANDBY Response Not Obtained 5.2 IF letdown has isolated due to a plant transient, THEN establish normal letdown using ATTACHMENT 1, RESTORING LETDOWN. 5.3 IF a letdown isolated due to a system malfunction, THEN perform the following: [] Attempt to restore any letdown flow using FNP-1-AOP-16.0, CVCS MALFUNCTION. [] Continue with applicable steps of this procedure. 5.4 WHEN normal letdown restored IF required, THEN return to step 4.4 to establish auxiliary spray. o Step 7 continued on next page Page 6 of 11 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.2 RCS Minimum Temperature for Criticality RCS Minimum Temperature for Criticality 3.4.2 LCO 3.4.2 Each RCS loop average temperature (Tavg) shall be 541°F. APPLICABILITY: MODE 1, MODE 2 with kelt 1.0. ACTIONS CONDITION

.---.. --..... REQUIRED ACTION Tavg in one or A.1 loops not within limit. /

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.4.2.1 Verify RCS Tavg in each loop 541°F. Farley Units 1 and 2 3.4.2-1 COMPLETION TIME FREQUENCY

N 0

Only required if low low T avg alarm not reset and any RCS loop Tavg < 547°F 30 minutes thereafter Amendment No. 146 (Unit 1) Amendment No. 137 (Unit 2)

4. 004G2.4.21 001/NEW/RO/C/A 4.0/4.6/N/N/2ICVRIY A LOCA and LOSP has occurred on Unit 1, and the following conditions occurred:

FRP-C.2, Response To Degraded Core Cooling, is in progress.
CCW to ALL the RCPs thermal barriers have been lost.
All charging pumps have tripped.
All RCP's are secured.
The five hottest CETCs are; 773°F, 779°F, 1023°F, 1252°F, 1508°F and all stable.
All SG pressures are at 1000 psig.
Off-Site Power is available.

Which one of the following states: 1) the FRP that must be in effect for the conditions given (FRP-C.2 Response To Degraded Core Cooling, OR FRP-C.1 Response To Inadequate Core Cooling), and 2) whether the RCPs will be started or not? A. Enter FRP-C.1 RCPs will be started B. Enter FRP-C.1 RCPs will NOT be started C. Remain in FRP-C.2 RCPs will be started Dy Remain in FRP-C.2 RCPs will NOT be started Page: 8 c:i 200 1211412009 A -Incorrect. The first part is incorrect, but is plausible. The fifth hottest core Core Exit Thermo Couple is not higher than 1200°F, so FRP-C.1 is not entered, but the highest three are >1200°F. Confusion may exist as to which of the highest thermocouples have to be >120QoF prior to entry into FRP-C.1. If the first part was correct, the second part would be correct also. A major difference between FRP-C.1 & C.2 is that in FRP-C.1, RCPs are started as a last resort even with no support conditions. In FRP-C.2 a RCP is started only if all support conditions are met. B -Incorrect. The first part is incorrect (see A). The second part is incorrect, but plausible, since for FRP-C.2 and all other procedures it is correct. Confusion may exist as to whether or not FRP-C.1 directs starting the RCPs without support conditions when FRP-C.2, and all other procedures do not. C -Incorrect. The first part is correct (see D). The second part is incorrect (see A). 0-Correct. The fifth hottest Core Exit TC is > 700°F but <1200°F. Therefore, FRP-C.2 is still in effect and FRP-C.1 is not entered. FRP-C.2 does not direct starting a RCP without support conditions, but FRP-C.1 does. CSF-O, Critical Safety Function Status Trees, Revision 17 FNP-1-FRP-C.1, Response To Inadequate Core Cooling, Revision 17 FNP-1-FRP-C.2, Response To Degraded Core Cooling, Revision 17 Page: 9 of 200 1211412009 Previous NRC exam history if any: 004G2.4.21 004 Chemi cal and Vol ume Control System 2.4.21 K ncmledge eX the parameters and logic used to assess the status eX safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. (CFR: 41.7/43.5/45.12) RO 4.0 SR04.6 Match justification: RO level knowledge of the entry condition parameters and logic used to assess Red and Orange path (and logic) of the Critical Safety Functions is required to answer this question correctly. The Charging pumps in the CVCS system (which are also HHSI pumps during a LOCA) are provided as tripped which prevent all RCP support conditions from being met (along with a loss of CCW to the RCP Thermal barriers which is also listed in the stem). The second part of the question directly addresses the effect of the CVCS system on the procedure directions concerning starting or not starting RCPs. Objective:

1. EVALUATE plant conditions and DETERM I NE if entry into (1) FRP-C.1, Responseto I nooequate Core Cool i ng; or (2) FRP-C.2, Response to Degrooed Core Cool i ng; or (3) FRP-C.3, Response to Saturated Core Cool i ng is requi red. (OPS-52533C02)

2. EVALUATE plant conditions and DETERM I NE if any system components need to be operated whi I e performi ng (1) FRP-C.1, Response to I nooequate Core Cool i ng; (2) FRP-C.2, Response to Degrooed Core Cool i ng; (3) FRP-C.3, Response to Saturated Core Cool i ng. (OPS-52533C06)

Page: 10 of 200 1211412009 8/2912007 08:33 FNP-1-CSF-O.2 o NO RCS SUBCOOLING FROM CORE EXIT TC'S GRTRTHAN 16 0 F {45° F} CORE COOLING FIFTH HOTTEST CORE EXIT TC LESS THAN 700 0 YES YES Page 1 of 1 .... Revision 17 GO TO FRP-C.3 CSF SAT FNP-1-FR RESPONSE TO INADEQUATE CORE COOLING Revision 17 Step Action/Expected Response n 19 Check core cooling. 19.1 Check core exit T/Cs -LESS THAN 1200°F. 19.2 Check at least two RCS hot leg temperatures -LESS THAN 350° F. RCS HOT LEG TEMP [] TR 413 19.3 Check REACTOR VESSEL LEVEL indication -GREATER THAN 0% UPPER PLENUM. 20 Go to FNP-1-EEP-1. LOSS OF REACTOR OR SECONDARY COOLANT. step 14. Response NOT Obtained 19.1 Proceed to Step 21. OBSERVE NOTE PRIOR TO STEP 21. 19.2 Return to step 17. 19.3 Return to step 17. NOTE: Normal ---------- conditions are desired but not required for starting RCPs. J ------------ =< 21 Check if RCPs should be I Il started. H \Ld 21.1 Check core exit TiCs -GREATER 21.1 Proceed to step 22. THAN 1200°F. Step 21 continued on next page. _Page Completed Page 22 of 33 FNP-l RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response Response NOT Obtained n ************************************************************************************** CAUTION: Further degradation of core cooling can occur if any running RCP is stopped before being directed by this procedure even if normal support conditions are lost. ************************************************************************************** 1 Monitor RWST level. 2 RWST LVL [] LI 4075A [] LI 4075B 1.1 rCA] WHEN RWST level less than 12.5 ft, THEN go to FNP-I-ESP-l.3, TRANSFER TO COLD LEG RECIRCULATION. Verify proper SI valve alignment using ATTACHMENT

2. SI VALVE ALIGNMENT FOR COLD LEG INJECTION.

___ Page Completed Page 2 of 22 / it.: \ \.1 V !. FNP-I-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step n 3 Action/Expected Response Check any HHSI flow -GREATER THAN 0 gpm. A TRN HHSI FLOW [] FI 943 HHSI B TRN RECIRC FLOW [] FI 940 3 Response NOT Obtained Perform the following. 3.1 Verify all available charging pumps started. CHG PUMP [] lA amps > 0 [] IB amps > 0 [] lC amps > 0 3.2 Verify charging pump MOV disconnects closed using ATTACHMENT 3, CHARGING PUMP MOV DISCONNECTS. 3.3 Verify proper SI alignment. CHG PUMPS TO REGENERATIVE HX [] QIE21MOV8107 closed [] QIE21MOV8108 closed RWST TO CHG PUMP [] QIE21LCVllSB open [] QIE21LCVllSD open VCT OUTLET ISO [] QIE21LCVllSC closed [] QIE21LCVllSE closed HHSI TO RCS CL ISO [] QIE21MOV8803A open [] QIE21MOV8803B open [ ] [] [ ] [ ] CHG PUMP SUCTION HDR ISO QIE21MOV8130A QIE21MOV8130B QIE21MOV8131A QIE21MOV8131B CHG PUMP DISCH HDR ISO open open open open [] QIE21MOV8132A open [] QIE21MOV8132B open [] QIE21MOV8133A open [] QIE21MOV8133B open Step 3 continued on next page. _Page Completed Page 3 of 22 l, .. c.' L\\I\'. !'. FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response Response NOT Obtained n _Page Completed 3.4 HHSI flow now established, THEN proceed to step 4, IF NOT, establish HHSI bypass SI flow. CHG PMP RECIRC TO RCS COLD LEGS [] Q1E21MOV8885 open HHSI TO RCS CL ISO [] Q1E21MOV8803A closed [] Q1E21MOV8803B closed 3.5 IF HHSI flow now established, THEN proceed to step 4, IF NOT, perform the following. 3.5.1 Open HHSI isolation valves. HHSI TO RCS CL ISO [] Q1E21MOV8803A [] Q1E21MOV8803B 3.5.2 Align charging pump suction header isolation valves based on 1B charging pump status. 1B Charging Pump Aligned As Aligned As Status A Train pump B Train pump CHG PUMP SUCTION HDR ISO Q1E21MOV [ ] 8130A [] 8130A open closed [ ] 8130B [ ] 8130B open closed [] 8131A [] 8131A closed open [] 8131B [] 8131B closed open Step 3 continued on next page. Page 4 of 22 FNP-1-FRP-C.2 Step n NOTE: " T\',i ';1__,:'--- _______ --,. ________ -. ,[ RESPONSE TO DEGRADED CORE COOLING Revision 17 Action/Expected Response Response NOT Obtained 3.5.3 Align charging pump discharge header isolation valves based on 1B charging pump status. 1B Charging Pump Aligned As Aligned As Status A Train pump B Train pump CHG PUMP DISCH HDR ISO Q1E21MOV [ ] 8132A [ ] 8132A open closed [] 8132B [ ] 8132B open closed [ ] 8133A [ ] 8133A closed open [] 8133B [ ] 8133B closed open Continuing efforts to establish SI flow should not interfere with performance of the remainder of this procedure. 3.6 If HHSI flow NOT established, THEN Continue efforts to establish SI flow.

HHSI flow I
LHSI flow I
Any form of RCS injection.

I _Page Completed Page 5 of 22 __________ -r __________________ .* ____________ __ -, ______________ -, :L;<,i):'.L .1 FNP-l-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response Response NOT Obtained n ************************************************************************************** CAUTION: Pump damage may occur if RHR pumps are operated on miniflow for longer than three hours with no CCW supplied to the RHR heat exchangers.

4 Check LHSI status. 4.1 Verify CCW flow to RHR heat exchangers -ESTABLISHED. CCW TO lA(lB) RHR HX [] QlP17MOV3l85A open [] QlP17MOV3l85B open 4.2 Check RCS pressure -LESS THAN 275 psig(435 psig}. lC (lA) LP RCS NR PRESS [] PI 402B [] PI 403B 4.2 Proceed to step 5. Step 4 continued on next page. ___ Page Completed Page 6 of 22 'c) I."!!. i .i FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step n Action/Expected Response 4.3 Check both RHR flows -GREATER THAN 1.Sx10 3 gpm. 1A(lB) RHR HDR FLOW [] FI 60SA [] FI 60SB _Page Completed Page 7 of 22 Response NOT Obtained 4.3 Verify LHSI properly aligned. RHR PMP [] 1A amps > 0 [] 1B amps > 0 1A(lB) RHR HX TO RCS COLD LEGS ISO [] Q1E11MOV8888A open [] Q1E11MOV8888B open RWST TO 1A(lB) RHR PUMP [ ] Q1EllMOV8809A open [] Q1EllMOV8809B open 1A(lB) RHR HX DISCH VLV [ ] HIK 603A open [ ] HIK 603B open 1A (lB) RHR HX BYP FLOW VLV [] FK 60SA closed [] FK 60SB closed 1A(lB) RHR TO RCS XC ON [] Q1EllMOV8887 A open [] Q1EllMOV8887B open HOT LEGS ".,) L\; "i FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response Response NOT Obtained n 5 Check RCS vent paths. 5.1 Check any PRZR PORV ISO -POWER AVAILABLE. 5.2 Verify both PRZR PORVs -CLOSED. 5.1 Restore power to PRZR PORV ISO valves unless de-energized for inoperable PORVs not capable of being manually cycled. 5.2 Perform the following. 5.2.1 Close PRZR PORVs. 5.2.2 IF any valve can NOT be closed. THEN close its PORV ISO valve. NOTE: The purpose of the following step is to establish an available PORV flowpath for mitigation of overpressure conditions. without relying on the PRZR code safety valves. A failed open PORV must not be unisolated. A leaking PORV which is isolated with power available to the isolation valve should remain isolated until needed to reduce RCS pressure or mitigate an RCS overpressure condition. Any leaking PORV should be re-isolated when not in use. 5.3 Check at least one PRZR PORV ISO -OPEN. 5.4 Verify reactor vessel head vent valves -CLOSED. RX VESSEL HEAD VENT OUTER ISO [] Q1B13SV2213A [] Q1B13SV2213B RX VESSEL HEAD VENT INNER ISO [] Q1B13SV2214A [] Q1B13SV2214B ___ Page Completed Page 8 of 22 5.3 upen any PRZR PORV ISO not required to isolate an open or leaking PORV. FNP-1-FRP-C.2 it) .ii.:' RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response Response NOT Obtained n 6 Check RCP status. 6.1 Check at least one RCP -STARTED. 6.1 Proceed to Step 8. ************************************************************************************** CAUTION: To prevent potential seal damage, neither seal injection nor CCW cooling should be restored to a RCP which has lost both seal injection and CCW cooling. ************************************************************************************** NOTE: Normal support conditions for running RCPs are desired, however, RCP operation must continue even if support conditions cannot be maintained. 6.2 Verify No.1 seal support conditions established. 6.2.1 rCA] Maintain seal injection flow -GREATER THAN 6 gpm. 6.2.2 Verify No.1 seal leakoff flow -WITHIN FIGURE 1 LIMITS. 6.2.3 Verify No.1 seal differential pressure -GREATER THAN 200 psid. 6.3 Verify CCW -ALIGNED. CCW FROM RCP THRM BARR [] Q1P17HV3045 open [] Q1P17HV3184 open ___ Page Completed Step 6 continued on next page. Page 9 of 22 !.\) .l v>: 1. : FNP-l-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response n I 6.4 Check RCP thermal barrier -INTACT. RCP THRM BARR CCW FLOW HI [ ] Annunciator DD2 clear 6.5 Check CCW to RCP oil coolers -SUFFICIENT. CCW FLOW FROM RCP OIL CLRS LO [ ] Annunciator DD3 clear 6.6 Check RCP oil level -SUFFICIENT. RCP lA(lB,lC) BRG UPPER/LOWER OIL RES LO LVL [ ] Annunciator HHl clear [ ] Annunciator HH2 clear [] Annunciator HH3 clear I Response NOT Obtained I I 6.4 Verify CCW flow isolated. CCW FROM RCP THRM BARR [] QlP17HV3045 closed [] QlP17HV3l84 closed 6.5 Verify CCW -ALIGNED. CCW TO RCP CLRS [] QlP17MOV3052 open CCW FROM RCP OIL CLRS [] QlP17MOV3046 open [] QlP17MOV3l82 open NOTE: Since RCP damage may occur when operating RCPs without normal support conditions established or under highly voided RCS conditions, the intent of the following step is to save one RCP (which provides the best pressurizer spray capability) for future use, if all three RCPs are running. 7 Check if one RCP should be stopped. 7.1 Check ALL RCPs -STARTED 7.2 Stop RCP lB. 7.3 Proceed to Step 9. _Page Completed Page 10 of 22 7.1 Proceed to Step 9.

\ ...... t**" ' E FNP-I-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response n 8 Check core cooling. S.l Check REACTOR VESSEL LEVEL indication

-GREATER THAN 0% UPPER PLENUM. S.2 Check core exit TiCs -LESS THAN 700°F. S.3 Go to procedure and step in effect. 9 Check SI accumulator discharge valve status. 9.1 Check power to discharge valves -AVAILABLE. lA(1B,lC) ACCUM DISCH ISO [] QIE21MOVSSOSA [] QIE21MOVSSOSB [] QIE21MOVSSOSC 9.2 Check discharge valves -OPEN. lA(1B,lC) ACCUM DISCH ISO [] QIE21MOVSSOSA [] QIE21MOVSSOSB [] QIE21MOVS808C 10 Monitor CST level. 10.1 rCA] Check CST level greater than 5.3 ft. CST LVL [] LI 4132A [] LI 4132B 10.2 Align makeup to the CST from water treatment plant OR demin water system using FNP-1-S0P-5.0, DEMINERALIZED MAKEUP WATER SYSTEM, as necessary. _Page Completed Page 11 of 22 Response NOT Obtained S.l IF SI established, THEN return to step 2, IF NOT, proceed to step 9. S.2 IF core exit TiCs falling, THEN return to step 2, IF NOT, proceed to step 9. 9.1 Close accumulator discharge valve disconnects using ATTACHMENT

1. 9.2 IF accumulators have NOT discharged, THEN open discharge valves. lA(1B,lC)

ACCUM DISCH ISO [] QIE21MOV8808A [] QIE21MOV8808B [] QIE21MOV8808C 10.1 Align AFW pumps suction to SW using FNP-I-S0P-22.0, AUXILIARY FEEDWATER SYSTEM.

\! .1.; .,. FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response Response NOT Obtained n **************************************************************************************

CAUTION: To prevent potential release of radioactive material to the atmosphere, a faulted or ruptured SG should only be used if no intact SG is available.

11 Check intact SG levels. 11.1 Check narrow range levels -GREATER THAN 31%{48%}. 11.1 Verify total AFW flow to intact SGs greater than 395 gpm. AFW FLOW TO 1A(1B,lC) SG [] FI 3229A [] FI 3229B [] FI 3229C AFW TOTAL FLOW [] FI 3229 Step 11 continued on next page. ___ Page Completed Page 12 of 22 ,l .... i i.':' FNP-I-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response Response NOT Obtained n 11.2 [CAl WHEN intact SG narrow range level 31%-65%{48%-65%}, THEN maintain intact SG narrow range level 31%-65%{48%-65%}. 11.2.1 Control MDAFWP flow. MDAFWP FCV 3227 RESET [] A TRN reset [] B TRN reset MDAFWP TO lA/IB/IC SG B TRN [] FCV 3227 in MOD Intact SG lA MDAFWP TO lA(lB,IC) SG QIN23HV [ ] 3227A [ ] in MOD in MDAFWP TO lA(lB,IC) SG FLOW CONT IB 3227B MOD HIC []3227AA [] 3227BA adjusted adjusted lC [ ] 3227C in MOD []3227CA adjusted Step 11 continued on next page. _Page Completed Page 13 of 22 \ .. ,J .i.e! I,!: i. FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response n I I 11.2.2 Control TDAFWP flow. TDAFWP FCV 3228 [] RESET reset TDAFWP SPEED CONT [ ] SIC 3405 adjusted Intact SG 1A 1B 1C TDAFWP TO 1A(1B,lC) SG Q1N23HV [] 3228A [] 3228B [] 3228C in MOD in MOD in MOD TDAFWP TO 1A(1B,lC) SG FLOW CONT HIC [] 3228AA [] 3228BA [] 3228CA adjusted adjusted adjusted _Page Completed Page 14 of 22 Response NOT Obtained I

FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response Response NOT Obtained n ************************************************************************************** CAUTION: Performance of step 12 will cause accumulator injection which may result in a red path on the INTEGRITY status tree. This procedure should be completed before transition to FNP-1-FRP-P.1. RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITIONS.

NOTE: After the low steam line pressure SI is blocked. excessive opening of steam dumps can cause a high steam flow LO-LO TAVG main steam isolation signal. 12 Reduce pressure in all intact SGs to 100 psig. 12.1 WHEN P-12 light lit (543°F). THEN perform the following. 12.1.1 Block low steam line pressure SI. STM LINE PRESS SI BLOCK -RESET [] A TRN to BLOCK [] B TRN to BLOCK 12.1.2 Verify blocked indication. BYP & PERMISSIVE STM LINE ISOL. SAFETY INJ. [] TRAIN A BLOCKED light lit [] TRAIN B BLOCKED light lit 12.1.3 Bypass the steam dump interlock. STM DUMP INTERLOCK [] A TRN to BYP INTLK [] B TRN to BYP INTLK Step 12 continued on next page. ___ Page Completed Page 15 of 22 !,; ,'<:v. ..:I FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step n Action/Expected Response 12.1.4 Adjust steam header pressure controller to control cooldown rate. STM HDR PRESS [] PK 464 adjusted 12.2 [CAl Maintain RCS cold leg cooldown rate -LESS THAN 100°F IN ANY 60 MINUTE PERIOD. 12.3 IF condenser available. THEN dump steam to condenser from intact SGs. BYP & PERMISSIVE COND AVAIL [] C-g status light lit STM DUMP [] MODE SEL A-B TRN in STM PRESS STM DUMP INTERLOCK [] A TRN in ON [] B TRN in ON STM HDR PRESS [] PK 464 adjusted Response NOT Obtained 12.3 Dump steam to atmosphere. 12.3.1 Direct counting room to perform FNP-0-CCP-64S. MAIN STEAM ABNORMAL ENVIRONMENTAL RELEASE. 12.3.2 IF normal air available. THEN control atmospheric relief valves to dump steam from intact SGs. IF NOT. dump steam using FNP-1-S0P-62.0. EMERGENCY AIR SYSTEM. 1A(lB.1C) MS ATMOS REL VLV [] PC 3371A adjusted [] PC 3371B adjusted [] PC 3371C adjusted 12.3.3 II no source of air available. THEN locally control SG atmospheric relief valves with handwheel to dump steam from intact SGs. (127 ft. AUX BLDG main steam valve room) Intact SG 1A 1B 1C Q1N11PCV [] 3371A [] 3371B [] 3371C 12.4 Check all intact SG pressures 12.4 Return to Step 11. OBSERVE -LESS THAN 100 psig. CAUTION PRIOR TO STEP 11. Step 12 continued on next page. _Page Completed Page 16 of 22 X) .l,\i X E :: FNP-I-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response n 12.5 Check at least two RCS hot leg temperatures -LESS THAN 350 0 F. RCS HOT LEG TEMP [] TR 413 12.6 Stop SG pressure reduction. STM HDR PRESS [] PK 464 adjusted lA(lB,lC) MS ATMOS REL VLV [] PC 3371A adjusted [] PC 3371B adjusted [] PC 3371C adjusted lA(lB,lC) MS ATMOS REL VLV [] QINIIPCV3371A closed [] QINIIPCV3371B closed [] QINI1PCV3371C closed Response NOT Obtained 12.5 Return to Step 11. OBSERVE CAUTION PRIOR TO STEP 11. ************************************************************************************** CAUTION: Pump damage may occur if RHR pumps are operated on miniflow for longer than 3 hours with no CCW supplied to the RHR heat exchangers.

13 Verify RHR pumps -STARTED. RHR PUMP [] 1A amps > 0 [] IB amps > 0 _Page Completed Page 17 of 22

I '\i pel, .t1 FNP-I-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response n 14 [CAl Check if SI accumulators should be isolated. Response NOT Obtained NOTE: Step 14.1 is a continuing action. 14.1 [CAl Check at least two RCS hot leg temperatures -LESS THAN 350°F. RCS HOT LEG TEMP [] TR 413 14.2 Reset SI. [] MLB-l 1-1 not lit (A TRN) [] MLB-l 11-1 not lit (B TRN) 14.3 Close all SI accumulator discharge valves. lA(lB,IC) ACCUM DISCH ISO [] QIE21MOV8808A [] QIE21MOV8808B [] QIE21MOV8808C _Page Completed 14.1 Perform the following. 14.1.1 WHEN at least two RCS hot leg temperatures are less than 350°F, THEN perform steps 14.2 and 14.3 to isolate accumulators. 14.1.2 Proceed to step 15. OBSERVE CAUTION PRIOR TO STEP 15. 14.2 IF any train will NOT reset using the MCB SI RESET pushbuttons, THEN place the affected train S821 RESET switch to RESET. (SSPS TEST CAB.) 14.3 Perform the following. 14.3.1 Vent any SI accumulator that cannot be isolated. ACCUM N2 VENT [] HIK 936 open SI ACCUM lA(lB,IC) ACCUM N2 SUPP IVT ISO QIE21HV 14.3.2 lA IB lC [] 8875A [] 8875B [] 8875C open open open an accumulator can NOT be isolated or vented, THEN consult the TSC staff to determine contingency actions. Page 18 of 22 t'z; 1 f. :1. FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response Response NOT Obtained n ************************************************************************************** CAUTION: Core cooling may degrade during subsequent steps. FNP-1-CSF-0.2. CORE COOLING status tree should be closely monitored.

15 Stop all Reps. RCP [] lA [] 1B [] 1C 16 Reduce pressure in all intact SGs to atmospheric pressure. 16.1 Maintain RCS cold leg cooldown rate -LESS THAN 100°F IN ANY 60 MINUTE PERIOD. 16.2 IF condenser available. THEN dump steam to condenser from intact SGs. BYP & PERMISSIVE COND AVAIL [] C-9 status light lit STM DUMP [] MODE SEL A-B TRN in STM PRESS STM DUMP INTERLOCK [] A TRN in ON [] B TRN in ON STM HDR PRESS [] PK 464 16.2 Dump steam to atmosphere. 16.2.1 Direct counting room to perform FNP-0-CCP-64S. MAIN STEAM ABNORMAL ENVIRONMENTAL RELEASE. 16.2.2 IF normal air available. [] [ ] [ ] THEN control atmospheric relief valves to dump steam from intact SGs. IF NOT. dump steam using FNP-1-S0P-62.0. EMERGENCY AIR SYSTEM. 1A(lB.1C) MS ATMOS REL VLV PC 3371A adjusted PC 3371B adjusted PC 3371C adjusted Step 16 continued on next page. _Page Completed Page 19 of 22 i./ }',! ,i ... ,\ FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step n 17 Action/Expected Response Verify any SI flow established.

Verify any HHSI flow -GREATER THAN 0 gpm. A TRN HHSI FLOW [] FI 943 HHSI B TRN RECIRC FLOW [] FI 940
Verify any LHSI flow -GREATER THAN 1.5x10 3 gpm. 1A(lB) RHR HDR FLOW [] FI 60SA [] FI 60SB _Page Completed Response NOT Obtained 16.2.3 IF no source of air available.

THEN locally control SG atmospheric relief valves with handwheel to dump steam from intact SGs. (127 ft. AUX BLDG main steam valve room) Intact SG 1A 1B 1C Q1NllPCV [] 3371A [] 3371B [] 3371C 17 Perform the following. 17.1 Continue efforts to establish SI flow.

HHSI flow
LHSI flow
Any form of RCS injection.

17.2 Return to Step 16. Page 20 of 22 FNP-I-FRP-C.2 t) I. ',i Ii j:v RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response Response NOT Obtained n 18 Check core cooling.

REACTOR VESSEL LEVEL indication

-GREATER THAN 0% UPPER PLENUM.

At least two RCS hot leg temperatures

-LESS THAN 350 0 F. RCS HOT LEG TEMP [] TR 413 19 Go to FNP-I-EEP-l. LOSS OF REACTOR OR SECONDARY COOLANT. step 14. -END-18 Page 21 of 22 Return to step 16 .

5. 004K1.04 OQ1/NE\iY/RO/C/A 3.4/3/8/N/N/2JCVRIY Unit 1 is a'CfbO%, and the following conditions occurred:

One Letdown orifice is on service.
LK-459F, PRZR LVL, controller demand has failed high. Which one of the following describes the effect on Charging Flow and RCP Seal Injection flows, with no operator actions? Charging Flow Seal injection Flows A'I Go up Go Down B. Go up Go up C. Go Down Go up D. Go Down Go Down Page: 11 eX 200 1211412009 A -Correct. When FK-122 fails high, charging flow increases.

This robs flow from the Seal injection lines and the Seal Injection flows go down. When Seal Injection flows go down, #2 seal flow and leakoff flow also goes down, since it is supplied by Seal Injection flow. When charging flow goes up, and letdown is unchanged, VCT level goes down. VCT pressure goes down due to expansion of the gas volume in the VCT. When pressure in the VCT goes down, #1 sealleakoff flow to the VCT goes up due to less back pressure. B -Incorrect. The first part is correct (see A). The second part is incorrect, due to the immediate effect of Seal Injection decreasing due to charging flow being in parallel with Seal inj. Flow. Charging flow increasing robs flow from seal injection flow. Plausible, since the VCT pressure goes down as VCT level goes down and the Number 1 seal leak off does go up eventually due to the VCT pressure drop, but the seal injection flow does not go up. C -Incorrect. Charging flow goes up due to the direct relationship between the master LK-459 Pressurizer level controller and the slave FK-122 controller, and the valve position of FCV-122. Plausible, since some of the MCB master slave controllers have an inverse relationship, such as PK-444A, PRZR PRESS REFERENCE controller and PK-444C & D, 1A & 1B LOOP SPRAY VLV controllers. The SPRAY VLV controller demands go up when the REFERENCE controller demand goes up. The second part is incorrect (see B). Plausible, since if the first part were correct, the second part would be correct also. D -Incorrect. The first part is incorrect (see C). The second part is correct (see A). Plausible, since physical connections and the cause/effect relationships between the CVCS system and the RCPS may be misunderstood and confusion could exist as to the inverse relationship between the two flows. FSD: CVCS'HHSI/A CCU MULA TORlRMWS A-181 009 PI D 175039 SH 6, eves chg & seal injection REACTOR COOLANT PUMPS, OPS-621 01 D, OPS-521 01 D, OPS-40301 D, STUDENT TEXT Page: 12 of 200 1211412009 Previous NRC exam history if any: 004K1.04 004 Chemi cal and Vol ume Control System K 1 Knowledge of the physical connections and/or cause-effect relationships between the eves and the follcming (CFR: 41.2 to 41.9 / 45.7 to 45.8) K1.04 RCPS, including seal injEdion flows ................................ 3.4 3.8 Match justification: The RCP #2 seal flow and the Seal Injection flows are both affected by the CVCS system during a Charging flow and/or VCT level/pressure transient. To correctly answer this question, knowledge of this relationship as well as the physical connections is required. Objective:

4. EVALUATE plant conditions and DETERM I NE if any system components need to be operated while performing AOP-100, Instrument Malfunction. (OPS-S2S21Q06).

2. RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the components associated with the Chemical and Volume Control System, to include the components found on Figure 3, Chemical and Vol ume Control System and Fi gure 4, RCP-Sea1 I nj ecti on System (OPS-40301 F02). Page: 13 of 200 1211412009 Date: 10/7/2009 tl I -175039 SN.1 8-12 j C'lCS CHARGING ........ 110E386 SN.1 LOC(i-1) o aV492A 0-175039 SH.l J-9 eveG !XAL INJ 110E86 SH.l LOC (a-9) -I DH H1E2lV136 J QV406C 3/4" 0 \35 1-2-176 /-£) r-----t5'<'\---'tIo:J---

A// @------T-----@;----J / 1 121 R P0142A HI REF. C--, ANNUN \ REf. G 3/4"-CCD-240 QV4978 3/'- I l-CS-2S01R CONNECTION QV494B 3/'-0 .... t--l 2-CS-2501R QV259A M 1-8109A 2_TM l 78FNW eHG PUMP BYPASS ORIFICE NO 1 ITEM ORCP FO 6004 (HOlE 4) !-PI ISlE _ eea-16 QV120A 1-8479A 3/4-n8 Time I 11 :08:41 AM aVIS7 1-8436 l-C78 3/4" V I" tiCB-\1 I I I I I I I j I---:' .... L._ I I I I I I I I I I I I I I I SIS RHRS HX 110[389 LOC G-6 OV263B \FlY 1-81168 'b--_l'--ZO-175037 SH.2(C ! 3/4-RV72SWB t Res PRT 110[.387 SH.2 lOC UI Title: C:\Reference Disk\Exam Reference Disk\Drawings\D17S039-0006.cal FROM eves CHG. PUMP DISCHARGE FROM REACTOR COOLANT PUMPS RCP SEAL STANDPIPE ___ __ _ ABOVE #3 SEAL 8539 REACTOR COOLANT IN Figure 10 -Rep -Seal Injection System FROM RCP B&C #1 SEAL BYPASS OpsRcp012 TO #3 :::;EA! I FAKOFE .. CTMT SUMP '-.l > 8121 (150PSIG) TO PRT OPS-62101D-52101D140301DIESP-52101D-Version 1 /"' '\ 6. 005A 1.07 001/NEyy/RO/C/A 2.5/3.1/N/N/3/CVRIY A time strole of Q1 E11 MOV8889, RHR TO RCS HOT LEGS ISO, in the open direction has been performed per STP-11.6, Residual Heat Removal Valves Inservice Test. Open direction ACCEPTABLE STROKE TIME RANGE is 9.96 to 13.47 Sec. Open direction MAXIMUM ALLOWABLE TIME is 16 Sec. Stroke times obtained were as follows:

At 1000 First time stroke:
At 1005 Second time stroke: 15.35 Secs 15.52 Secs Which one of the following describes MOV-8889 OPERABILITY lAW Technical Specifications and what the CR will require for these results lAW STP-11.6?

A'I* MOV-8889 is OPERABLE

Analysis of the time stroke results within 96 hours to determine if new stroke time is acceptable.

B.* Declare MOV-8889 INOPERABLE

Analysis of the time stroke results within 96 hours to determine if new stroke time is acceptable.

C.* MOV-8889 is OPERABLE

Repair or replacement of MOV8889. D.* Declare MOV-8889 INOPERABLE
Repair or replacement of MOV8889. A -Correct. Tech Specs requires the time stroke to be less than the Maximum stroke time. This is stated as acceptance criteria in STP-11.6, Step 5.3.3.4 & 5.3.3.5, but outside of the Acceptable Stroke Time Range the valve must have a retest and an analysis if the stroke time is still outside of the Acceptable range but less than the maximum. The second part is correct per STP-11.6, Step 5.4.2. B -Incorrect.

The first part is incorrect, but plausible. The tech spec limit is the same as the maximum time for the valve stroke. A stroke time above the maximum does not meet acceptance criteria and requires declaring the valve inoperable, but above the acceptable range AND below the Max time meets acceptance criteria. Acceptable range may be confused with acceptance criteria. Not meeting acceptance criteria indicates inoperability due to TS requirements not being met. Also, if either of the tests were greater than the maximum, or if no retest was possible this choice would be correct. Second part is correct (see A). C -Incorrect. First part is correct. Second part is incorrect but plausible. Writing a CR is required, but requiring repair or replacement of the valve is only required for a Page: 14 of 200 1211412009 valve that has been required to be declared inoperable (no analysis in 96 hours, greater than MAX stroke time, or no retest possible and outside of the acceptable range). Analysis and possible resetting the baseline of the valve stroke time is allowed and required by the STP. D -Incorrect. First part is incorrect (see B). Second part is incorrect (see C) but plausible. If the first part was correct, then this would be correct per STP-11.6, AND tech specs would not be met until the valve was repaired to allow time stroking in less than the Max allowed time. STP-11.6 step 5.4 Version 36 5.4 In Table 1, compare Actual Stroke Times to Maximum Allowable Times and to the Acceptabl e Stroke Ti me Range and perform the foil owi ng as appl i cabl e: Page: 15 of 200 5.4.1 IF the Actual Stroke Time for a valve exceeds the Maximum Allowable Time, THEN perform the following:

1. Decl are the val ve i noperabl e. 2. Check the appropri ate T echni cal Speci fi cati ons, T echni cal Requi rements Manual, and Fourth 10-Year Interval 1ST Program for corrective action requi rements. 5.4.2 I F the Actual Stroke Ti me for a valve is outside the Acceptable Stroke Ti me Range AND does NOT exceed the Maximum Allowable Time, THEN perform the following:

1. I mmedi ately retest the valve. 2. IF it is NOT possible to retest the valve, THEN declare the valve inoperable.

3. I F the val ve is retested AND the second set of data is aI so outsi de the Acceptable Stroke Time Range, THEN perform the following:

a. Submit a CR to have the data analyzed withi n 96 hours to verify that the new stroke time represents acceptable valve operation.

b. Enter the CR number in T abl e 1. c. Initiate an Admin LCO to declare the valve inoperable if not analyzed within 96 hours. 4. IF the valve is retested AND the second set of data iswithin the Acceptable Stroke Time Range, THEN analyze the cause of the initial deviation and submit a CR to have the resul ts documented in the Record of Tests. 5.4.3 IF any valve is declared inoperable, THEN perform the followi ng: 1. Resolve the unacceptable condition by performing one of the following:

-Repai r the val ve. -Replace the valve. -Anal yze the associ ated val ve stroke data to determi ne the cause of the devi ati on and whether valve operation is acceptable as is. 2. Prior to returning any valve to service following repair, replacement, or analysis, write a CR to request that ES issue new baseline data. 1211412009 Previous NRC exam history if any: 005A1.07 005 Residual Heat Removal System A1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design aSBX:iated with operating the RH RS controls induding: (CFR: 41.5/45.5) A 1.07 Determina:ion of test a::ceptcbility by comparison of rocorda::l valve response times with Te:::h-Spe::: r&:luiranents .............. 2.53.1* Match justification: Recorded values of valve response times are given and the applicant is required to assess whether or not Tech Specs are met on the RO level of knowledge. The STPs are the mechanism with which ROs assess operability of valves per their stroke times. The "Tech-Spec requirements" are assessed in the valve stroke STPs with acceptance criteria being met or not met. This question provides a stroke time with a retest (directed by the procedure in this case) and the applicant must assess "Tech-Spec requirements" as to declaring inoperable or not (in the first part of the answers), and further actions per the STP (in the second part of the answers). Objective: 1 RECALL AND APPLY the LCO and APPLICABILITY for Technical Specifications (TS) or TRM requirements, and the REQUI RED ACTIONS for 1 HR or less TS or TRM requi rements, and the rei evant porti ons of BA SES that 0 EFI N E the OPERA B I L I TY and APPLICABILITY of the LCO associated with the Residual Heat Removal System components and attendant equipment alignment, to include the following (OPS-52101 K01):

3.4.3, RCS Pressure and Temperature (PIT) Limits
3.4.6, RCS Loops-MODE 4
3.4.7, RCS Loops-MODE 5, Loops Filled
3.4.8, RCS Loops-MODE 5, Loops Not Filled
3.4.12, Low Temperature Overpressure Protection (L TOP) System
3.4.14, RCS Pressure Isolation Valve (PIV) Leakage
3.5.2, ECCS -Operati ng
3.5.3, ECCS -Shutdown
3.9.4, Residual Heat Removal (RHR) and Coolant Circulation

-High Water Level

3.9.5, Residual Heat Removal (RHR) and Coolant Circulation

-Low Water Level

13.5.1, Emergency Core Cool i ng System (ECCS) Page: 16 of 200 1211412009 05112/09 12:38:23 FNP-I-STP-ll.6 5.3 IF the RHR TO RCS HOT LEGS ISO, QIEIIMOV8889 Exercise Test is required, THEN perform the following:

5.3.1 Open the following valves: 5.3.1.1 5.3.1.2 RHR TO RCS HOT LEGS HDR TEST CONN ROOT, QIEll V048A. RHR TO RCS HOT LEGS HDR TEST CONN ROOT, QIEll V048E. 5.3.2 Record the pressure on Nl EIIPI2262. PRESSURE ______ PSI. 5.3.3 IF the pressure indicated on NIEI1PI2262 is less than OR equal to 550 psig, THEN perform the following: 5.3.3.1 IF required to isolate RHR TO RCS HOT LEGS ISO, QIEIIMOV8889 from an operating RHR train, THEN close RHR TO RCS HOT LEGS XCONN, QIEIIMOV8887A. 5.3.3.2 IF required to isolate RHR TO RCS HOT LEGS ISO, QIEI1MOV8889 from an operating RHR train, THEN close RHR TO RCS HOT LEGS XCONN, QIEIIMOV8887B. 5.3.3.3 Unlock and close disconnect switch QIRI8B036-B.

5.3.3.4 TO RCS HOT LEGS ISO QIEI time required for valve opening STROKE TIME column of Table I. LEPTANCE CRITERIA:

Stroke times are less than or equal to Maximum Allowable Times listed in Table 1.

5.3.3.5 RHR TO RCS HOT LEGS ISO QIEll a record time required for valve closing in STROKE TIME column of Table 1. '\ ( ACCEPTANCE CRITERIA:

Stroke times are less than or equal to the Maximum ) .. Allowable Times listed in Table 1. I Version 35.0 05112/09 12:38:23 FNP-I-STP-ll.6 5.4 In Table 1, compare Actual Stroke Times to Maximum Allowable Times and to the Acceptable Stroke Time Range and perform the following as applicable: 5.4.1 IF the Actual Stroke Time for a valve exceeds the 5.4.2 Allowable Time, THEN perform the following:

1. 2. Declare the Check the appropriate Technical Specifications, Technical Requirements Manual, and Fourth 1 0-Year Interval 1ST Program for corrective action requirements. IF the Actual Stroke Time for a valve is outside the Acceptable Stroke Time Range AND does NOT exceed the Maximum Allowable Time, _ THEN perform the following:

1. Immediately retest the A ,,+ __ ,;i)2.

THEN " { \" valve ______________ -__ . I AND the second of i; al;;-----' Ei--outside the Acceptable Stroke Time Range, THEN perform the /. 4" following: "ftcJ/{3> a. Submit a CR to have the within 96 hours to -? __ valve 5.4.3 operatIOn. --b. Enter the CR number in Table 1. c. Initiate an Admin LCO to declare the valve inoperable if not analyzed within 96 hours. 4. IF the valve is retested AND the second set of data is within the Acceptable Stroke Time Range, THEN analyze the cause of the initial deviation and submit a CR to have the results documented in the Record of Tests. IF any valve is declared inoperable, THEN perf.Qrm the following: ,..------=::=,;;;;"""'= . ...----:::=.

. .

unacceptable b)-'1>erforming o owmg: "'--* Repair the valve. ______

Replace the valve. ___ --r------___
Analyze the associated valve stroke data to determine the cause of the deviation and whether valve operation is acceptable as is. 2. Prior to returning any valve to service following repair, replacement, or analysis, write a CR to request that ES issue new baseline data. Version 35.0

/*******1

7. 005K4.0300;VNE IRO/C/A 2.9/3.21N/N/4/CVRIY Unit 1 is p'elforming a plant cooldown using the A Train RHR system, and the following conditions occurred:

HIK-603A, 1A RHR HX DISCH VLV, controller demand is at 50%.
FK-605A, 1A RHR HX BYP FLOW, controller is in AUTO with demand at 50%.
RHR flow on FI-605A, 1A RHR HDR FLOW, is 3100 gpm. At 1000:
HIK-603A demand setting is increased to 60%. At 1005:
RHR system flow is stable. At 1010:
Instrument Air is lost to FCV-605A due to an air supply line break. Which one of the following describes, with no operator actions: 1) RH R flow indicated on FI-605A at 1005, and 2) the position of FCV-605A at 101 O? At 1005 At1010 FI-605A, RHR HDR FLOW FCV-605A, 1A RHR HX BYP FLOW A. > 3100 gpm Closed B. 3100 gpm Open C. > 3100 gpm Open 3100 gpm Closed Page: 17 ci 200 1211412009 A -Incorrect.

This first part is incorrect, since even thought the flow does initially go up, the FT senses this and the HX BYP FK demands the HX BYP FCV to close down to maintain the 3100 gpm initial flow. Plausible, since flow does go up initially. Also, if the BYP FCV is in manual which it normally is, this choice would be correct. The second part is correct (see D). B -Incorrect. The first part is correct (see D). The second part is incorrect, but plausible. The valve fails closed on loss of air to maximize flow through the HX during a LOCA, but this valve could be confused with the HX discharge valve which fails open for the same reason. C -Incorrect. The first part is incorrect (see A). The second part is incorrect (see B). D -Correct. The design for the RHR HX BYP FCV is to operate in auto to maintain the total system flow rate constant while flow through the HX is adjusted with the potentiometer for the HX DISCH valve. The fail position of the valve is closed. FSD: A181002, Residual Heat Removal-LON Head Safety I njection Functional System De&:ription 3.15 RHR HEAT EXCHANGER DISCHARGE VALVES 5.1 RHR HEAT EXCHANGER BYPASS FLOW CONTROL Page: 18 of 200 1211412009 Previous NRC exam history if any: 005K4.03 005 Residual Heat Removal System K4 Knowledge cl RH RS design feature(s) and/or interlock(s) which prOllide or the following: (CFR: 41.7) K4.03 RHR he:t exchooger bypassflow control ............................ 2.93.2 Match justification: The design features of normal cooldown operation of the RHR HX BYP FCV (in auto adjusting to maintain constant total flow vice adjusting to maintain constant valve position-first part of choices) and the design fail position of the valve (closed-second part of choices) must be understood to correctly answer both parts of this question. Objective:

7. DEFINE AND EVALUATE the operational implications of normal I abnormal plant or Equipment conditions associated with the safe operation of the Residual Heat Removal System components and Equipment, to includethefollowing (OPS-40301K07):

Page: 19 cl 200

Normal Control Methods
A bnormal and Emergency Control Methods (Changes in system flow rates, Loss of control from the control room)
Automatic actuation including setpoints (examples-Reactor Trip, SI, PhaseA, LOSP/Iossof all AC power)
Protecti ve i sol ati ons such as hi gh flow, low pressure, low level i ncl udi ng setpoi nt
Protective interlocks
Actions needed to mitigate the consequence of the abnormal ity 1211412009 FNP Units 1 & 2 RESIDUAL HEAT REMOVAL A-181002 5.0 NON-CRITICAL COMPONENT FUNCTIONAL DESIGN REQUIREMENTS This section presents the functional design requirements for those components which are not critical to system function as defined in Section 3.0. These components enhance system performance, facilitate system maintenance or provide additional redundancy to components discussed in Section 3.0. 5.1 RHR HEAT EXCHANGER BYPASS FLOW

-Valves: l-RHR-FCV-60SB (QIEIIV033B) 2-RHR-FCV-60SA (Q2Ell V033A) 2-RHR-FCV-60SB (Q2Ell V033B) Flow Transmitters: QIEIIFT60SA QIEIIFT60SB Q2EIIFT60SA Q2EIIFT60SB 5.1.1 Basic Functions 5.1.2 Functional Requirements 5.1.2.1 5.1.2.2 5.1.2.3 5.1.2.4 The valves must be designed for pressure and temperature conditions of 600 psig and 400°F. (References 6.4.12, 6.S.S, 6.S.6 and 6.S.13) Valve thermal design transient requirements are summarized in Table T-14. (References 6.S.S and 6.S.6) Maximum allowable valve Cv equals 10S0 at 600 full open. (Reference 6.4.12) The design stroke time for opening or closing this operated valve is less than or equal to 10 seconds. S-1 Rev. IS I

8. 006K6.13 o01/IT!Sie 2.8/3.1/N/N/4/CVRIY A Small BreaKTOCA has occurred on Unit 1, and the following conditions occurred:

A reactor trip and safety injection is in progress.
1 A Charging Pump failed to auto start.
1 C Charging Pump is the only charging pump running.
RCS pressure is 1000 psig. Which one of the following states the Safety Injection flow indication on FI-943, A TRN HHSI FLOW, with no operator action? Safety Injection flow is approximately

___ _ A. Ogpm B. 150 gpm Cy 450 gpm D. 800 gpm Page: 20 of 200 1211412009 A -Incorrect. See C. Plausible since the meter is labeled "A train", and during cold leg recirc this meter indicates only A train flow which is 0 gpm with no A train charging pump running. However, the trains are cross connected during the injection phase, and the B train pump flow is also indicated by this meter during the injection phase. Normal Charging flow indicated by FCV-122 indicates 0 for this condition. B -Incorrect. See C. Plausible, since this is the approximate flow at normal RCS pressure with one charging pump. Also, it is the maximum charging indicated flow through the normal charging flow path at normal RCS pressure, but the normal charging flow path is isolated by the SI signal. Since the RCS pressure is less than NOP, the flow is greater than 150 gpm. C -Correct. At -1000 psig RCS Pressure, one HHSI (Charging) pump can provide about 450 gpm of flow. [Verified on simulator laptop, IC-73 SBLOCA from 100% power, 10,000 gpm leak. With RCS pressure at 1013 psig and one HHSI Pump tripped, HHSI flow on FI-943 was 440 gpm]. A knowledge of the exact value of charging flow from one pump at an RCS pressure of 1000 psig is not required to answer this question correctly. A knowledge of the characteristic pump curve for a centrifiugal pump and Charging pump capacity/capability at minimum is required. Also, knowledge of the system configuration in the injection phase of the LOCA (cross connected trains and both trains flow past the "A train" flow indicator). 3000 2000 1500 1000 500 0 -500 0 graph shows single pump curve, parrellel pump (2 pumps) curve, and a generic System characteristic curves for SBLOCA and LBLOCA. o -Incorrect. See C. Plausible, Two charging pumps could deliver 800 gpm into the RCS during a Large break LOCA if the RCS was at minimum pressure. This value is that which might be generally recalled from simulator observations for different conditions .. Page: 21 of 200 1211412009 Previous NRC exam history if any: 006K6.13 006 Emergency Core Cool i ng System K6 Knowledge of the effect of a loesor malfunction on the following will have on the ECCS: (CFR: 41.7/45.7) K6.13 Pumps ......................................................... 2.83.1 Match justification: This question requires knowledge of the effect on the ECCS system flow rate with one pump (HHSI) tripped. Objective:

2. RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the components associated with the Emergency Core Cooling System, to include the components found on Figure 2, Accumulators, Figure 3, Refueling Water Storage Tank, and Figure4, Emergency Core Cooling System (OPS-40302C02) . Page: 22 of 200 1211412009

9. 007A3.01 001/FNP BANK/RO/C/A 2.7/2.9/N/N/3/CVRIY Unit 2 is in Mode 5, and the following conditions occurred:

PI-472, PRT PRESS, reads 7.5 psig and is rising slowly.
LI-470, PRT LVL, reads 78% and is rising slowly.
RCS Pressure is 225 psig.
A Train RHR is aligned in the RCS Cooldown Operation lAW SOP-7.0, Residual Heat Removal System.
It has been determined that V8708A, A Train RHR Pump suction relief valve, is leaking by the seat. Which one of the following correctly states the impact on the PRT with no operator action and the required procedure actions to mitigate this condition per SOP-1.2, Reactor Coolant Pressure Relief System? The PRT Pressure will reach a maximum pressure of (1) psig, and to prevent reaching the PRT maximum pressure, the operator will be directed to (2) per SOP-1.2, Reactor Coolant Pressure Relief System. A. (1) 150 psig. (2) pump down the PRT with the RCDT pump and vent the PRT to #7 WGDT, if necessary.

B. (1) 150 psig. (2) gravity drain the PRT to the WHT. C. (1) 100 psig. (2) gravity drain the PRT to the WHT. (1) 100 psig. (2) pump down the PRT with the RCDT pump and vent the PRT to #7 WGDT, if necessary. Page: 23 c:I 200 1211412009 A -Incorrect. Part 1 incorrect, but plausible, since RCDT relief is set at 150 psig. Part 2 is correct (see D). B -Incorrect. Part one is incorrect (see A). Part 2 is incorrect, but plausible, since it would be correct IF the RCDT pumps were inoperable per SOP-1.2 step 4.3.3. Venting should not be necessary in this case due to the low energy of the RCS in mode 5 <<200°F), but the procedure does not address lowering pressure by just lowering level. Pressure is high because of level only, lowering level will also lower pressure and could be preferred, but lowering level by gravity draining should only be used if the RCDT pumps are inoperable. C -Incorrect. Part 1 is correct (see D). Part 2 is incorrect (see B). D -Correct. Both parts correct. RHR pump suction pressure is approx. the same as RCS pressure in this lineup: 225 psig per the stem. This makes it credible in that it could actually cause the rupture disc to break at it's setpoint of 100 psig per SOP-1.2 Step 3.5 "PRT pressure should be maintained < 100 psig to prevent rupture disc blowout." The PRT has a N2 pressure established of approx. 0.5 to 3 psig to prevent formation of explosive gasses. This bubble will compress as level rises. FNP-2-S0P-1.2, REACTOR COOLANT PRESSURE RELIEF SYSTEM, Version 30.0 4.4 Reducing PRT Pressure 4.4.1 Have Chemistry verify gas addition to the shutdown gas decay tank to be used for PRT venting (#7 or #8) is acceptable (e.g. H2 < 4% and 02 < 1% per CCP-203). RCDT Relief valve pressure is 150 psig Per U259507. PRT Rupture disks blows at 100 psig per SOP-1.2 STEP 3.5. Per SOP-1.2: 4.3.3 Gravity Draining PRT to WHT NOTES:

This method of draining the PRT should only be used if RCDT pumps are inoperable.

2-S0P-7.0, Residual Heat Removal System, Version 79.0 2-S0P-1.2, Reactor Coolant Pressure Relief System, Version 31 3.4 PRT level should be 68-78% during normal operation. 3.5 PRT pressure shoul d be mai ntai ned < 100 psi g to prevent rupture di sc blowout. 4.3.2 Draining the PRT Using an RCDT Pump [Normal preferred method] 4.3.3 Gravity Drai ni ng PRT to WHT NOTES:* This method of drainingthePRT should only be used if RCDT pumpsare inoperable. Page: 24 ci 200 1211412009 Previous NRC exam history if any: n/a 007A3.01 007 Pressurizer Relief Tank / Quench Tank System A3 Ability to monitor automatic operation of the PRTS, induding: (CFR: 41.7/45.5) A3.01 Components which diochcrgetothePRT ............................. 2.7* 2.9 Match justification: This question requires monitoring MCB PRT pressure indication and to know the pressure for automatic operation of the PRTS (at 100 psig the rupture disks ruptures). In this question, a component is discharging into the PRT (RHR Suction relief), and to answer this question, knowledge is required of what pressure will be indicated on the MCB prior to the PRT rupture disk automatically rupturing to relieve the pressure. Objective:

2. RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the components associated with the Pressurizer System, to include the components found on Figure 3, Pressurizer and Pressurizer Relief Tank (OPS-40301E02).

Page: 25 of 200 1211412009 07/02/096:42:21 FNP-2-S0P-l.2 3.0 Precautions and Limitations 3.1 PRT temperature should not exceed 120°F during normal plant operation. 3.2 PRT nitrogen overpressure of 0.5 to 3 psig should be maintained to prevent formation of an explosive hydrogen -oxygen mixture. PR T pressure should not exceed 6 psig during normal plant operation. PRT level should be 68-78% during normal operation. PRT pressure should < 100 psig to prevent rupture disc At least one of the two reactor vessel head vent system paths, vv",.,,,,,,u valves in series powered from the Auxiliary Building DC Distribution System, shall be OPERABLE and closed at all times when in Modes 1-4. (TR 13.4.3) 3.7 While stroking the upstream valve (Q2B13SV2214A or Q2B13SV2214B), MCB closed indication could be momentarily lost on the downstream valve (Q2B13SV2213A or Q2B13SV2213B) due to minor water hammer. This phenomenon is common and documented for Plant Farley and for other plants, and has been evaluated to have no detrimental impact. {CR 2007103114 } 3.8 While stroking the downstream valve (Q2BI3SV2213A or Q2B13SV2213B), MCB closed indication could be momentarily lost on the upstream valve (Q2B13SV2214A or Q2B13SV2214B) due to rapid depressurization across the upstream valve. This phenomenon is documented for Plant Farley and has been evaluated to have no detrimental impact. {CR 2007103114 } Version 30.0 07/02/09 6:42:21 FNP-2-S0P-1.2 4.3.2.18 IF RCDTwas aligned to WHT, THEN perform the following:

1. Close RCDT DISCH TO WHT, Q2G21 V009 (2-LWP-V-7137).

2. Open RCDT PUMP DISCH TO RHT ISO, Q2E21V315 (2 CVC V 8551). 4.3.2.19 Verify closed the following:

PRT N2 SUPPLY ISO Q2B31HV8047
PRT N2 SUPPLY ISO Q2B31HV8033
Nitrogen supply from bulk storage to PRT valve 2-GWD-V-7920 (Q2G22V215)

(121' PPR) 4.3.2.20 Align RCDT system as desired per FNP-2-S0P-50.0, LIQUID WASTE PROCESSING SYSTEM. 4.3.3 Gravity Draining PRT to NOTES:

This method of draining the PRT should only be used if ReDT pumps are inoperable.

__

The bottom of the PRT sparger is 12" =

gallons = The sparger is a 12" perforated pipe that sits 12" off the bottom of the PRT. The top of the sparger is at 24" = gallons = The level doesn't have to be below the bottom of the sparger because the pipe is perforated on all sides, but it may be desirable. 4.3.3.1 Verify closed the following:

PRT vent to GDT 2-RC-V-8025 (Q2B13V064), 121'.
PRT vent to SID Gas Decay Tanks 2-GWD-V-7935 (Q2G22V237),83' 4.3.3.2 Verify closed nitrogen/hydrogen supply to SID GDT's isolation valve 2-GWD-V -7849 (Q2G22V040).

4.3.3.3 Open nitrogen supply from bulk storage to PRT valve 2-GWD-V-7920 (Q2G22V215). 4.3.3.4 Verify PRT regulator 2-RC-PCV-8034 (Q2B13V042) adjusted to 3 psig. 4.3.3.5 Open the following PRT N2 SUPPLY ISO valves (MCB):

Q2B31HV8047
Q2B31HV8033 Version 30.0

10. 007EK2.02 001/NEW/RO/C/A 2.6/2.8/N/N/4/CVRIY Unit 1 was at 100% power, and the following conditions occurred:

FRP-S.1, Response To Nuclear Power Generation

-ATWT, is in progress.

The Main Turbine was unable to be tripped from the MCB.
A Safety Injection (SI) has NOT occurred.
Tavg is 563°F. Which one of the following describes the immediate effects if the Reactor Trip Breakers are opened locally at this time? A.. The Block of an Auto SI will be allowed.
The Feed Water Reg Valves will trip closed. B.* The Block of an Auto SI will be allowed.
The Steam Flow high setpoint will be reset. C.* The Main Turbine will trip.
The Feed Water Reg Valves will trip closed. D'!"'* The Main Turbine will trip.
The Steam Flow high setpoint will be reset. Page: 26 of 200 1211412009 A -Incorrect.

The first part is incorrect, since a block of SI is not an effect of opening the RT bkrs unless the SI has already initiated. Plausible, since if an SI had initiated this would be correct, and in many cases with an ATWT and NO Turbine trip an SI occurs, but the stem states that an SI has NOT occurred. The second part is incorrect also, since The Feed water Regulating Valves are only tripped closed by opening RT bkrs (P-4) in coincidence with a Low Tavg signal of 554°F. Since Tavg is still above 554°F, a FWIS will not occur immediately. Plausible, since on most reactor trips, a Low Tavg occurs due to steam dump operation very quickly after the Trip, but in this case Tavg is high due to the ATWT. B -Incorrect. The first part is incorrect (see A). The second part is correct (see D). C -Incorrect. The first part is correct (see D). The second part is incorrect (see A). o -Correct. P-4 will trip the Main Turbine regardless of any other plant condition or parameter(s), and this will occur immediately when the Reactor Trip breakers are open. The Steam Flow high setpoint will be reset by P-4 immediately when the Reactor trip breakers are open regardless of any other plant parameter (the actuation of the steam flow MSIV isolation signal requires a Low Low Tavg: P-12, but resetting the signal occurs regardless of Tavg on a reactor trip. These are two of the several functions of P-4. The MCB handswitch which trips the Turbine directly did not work in this scenario (it operates the 20 AST-2 relay). P-4 operates the 20AST-1 and 20ET relays which open the interface valve and bleed EH fluid off of Throttle valves & Reheat Stop valves to trip main turbine per Figure 19 in the Student text for Main Turbine. FNP-0-SOP-0.3, OPERA TI ONS REFERENCE I NFORMATI ON, APPENDI X G, OPERATIONAL PERMISSIVESAND CONTROL INTERLOCKS, Version 39.0 Permissive

1. P-4 Reactor Trip Interlock Source Reactor Trip and Bypass Breakers Setpoint Breakers Open Coincidence

& LightStatus RT A & BY A Open or RTB & BYB Open No Light Function Prevents a rapi d cool down of pri ma-y system after a reactor tri p. 1. Trips Turbi ne 2. Trips F.W. Reg Valves on Low Tavg 3. Seals in F.W. Reg Valve Tri ps from S.I. and S/G Hi Hi Level 4. Allows S.I. signal to be blocked after S.I. initiation

5. Resets Hi Stm Flow Setpoint 6. Arms steam dump system, enables plant trip controller and di sabl es loss of load control I er. Page: 27 cJ 200 1211412009 007EK2.02 007 Reactor Trip EK2 KnOlNleclgeof the interrelations between a reactor trip and thefollcming: (CFR 41.7 / 45.7) EK2.02 Bree:i<ers, relays end dis::onnocts

................................... 2.62.8 Match justification: To answer this question the applicant must know the normal relationship between the P-4 interlock and the reactor trip breakers and the various functions it accomplishes during a reactor trip. Objective:

1. RECALL AND DESCRI BE the operation and function of the following reactor trip signals, permissives, control interlocks, and engineered safeguards actuation signals associated with the Reactor Protection System (RPS) and Engineered Safeguards Features (ESF) to i ncl ude setpoi nt, coi ncidence, rate functi ons (if any), reset features, and the potential consequences for improper conditions to incl ude those items in the followi ng tables (OPS-52201107):

Table 1, Reactor Trip Signals
T abl e 2, Engi neered Safeguards Features A ctuati on Si gnal s
Table 5, Permissives
Table 6, Control interlocks

5. DEFI NE AND EVALUATE the operational impl ications of abnormal plant or equipment conditions associated with the operation of the Reactor Protection System (RPS) components and equi pment to i ncl ude the foil owi ng (OPS-522011 09). Page: 28 of 200

Normal Control Methods
Abnormal and Emergency Control Methods
Automatic actuation including setpoint ( example SI, PhaseA, Phase B, MSLIAS, L asp, SG I eve!)
A cti ons needed to mi ti gate the consequence of the abnormal i ty 1211412009 04/03/09 13: 17 :30 Permissive Source ReactorTrip

{I. P-4 Reactor* \ and Bypass Trip Interlock \ Breakers 2. P-6 I1R Power NIS 35 and 36 Escalation Permissive APPENDIXG FNP-O-SOP-O.3 APPENDIXG OPERATIONAL PERMISSIVES AND CONTROL INTERLOCKS Setpoint Breakers Open 10-10 amps PERMISSIVES Coincidence & Light Status Function / I RTA & BY A Open or Prevents a rapid cooldown of primary systemlafter a reactor trip. f\ RTB & BYB Open / / No Light f}-c 3. Seals in F.W. Reg Valve Trips from S.1. and S/G Hi Hi Level I ;7 112 > Setpoint Lit> Setpoint Permission to Block Source Range ,3 .. Allows S.1. signal to be blocke¥fter S.1. S. 5. Resets Hi Stm

6. Arms steam dump system, enables plant trip controller and disables loss of load controller.

Allows power escalation into the IR by turning Both Train A & B Source Range Block switches to Block. Above setpoint 1. Blocks SR Hi 0 Reactor Trip 2. Turns off Hi volt to SR Instr. Below setpoint Auto reinstates Hi volt to SR lnstr. Page 1 of7 Version 38.0

11. OOBAA2.26 001/NEW/RO/C/A 3.1/3.4/N/N/2ICVR/Y Unit 1 has manually initiated a Safety Injection due to rapidly falling pressurizer pressure, and the following conditions occurred:

At 1000:

Pressurizer level 35% and rising.
RCS pressure 1700 psig and falling.
PRT level is 73% and pressure is 5 psig and stable.
TI-453, PORV downstream temperature, is 11 rF.
TI-453, Safety Valve downstream temperature, is 101°F.
TI-453, Safety Valve downstream temperature, is 101°F.
TI-453, Safety Valve downstream temperature, is 102°F.
Containment Pressure 0.2 psig and slowly rising.
R-2, 7, 11 and 12 are in alarm. At 1015: Transition is made to EEP-1.0, Loss of Reactor or Secondary Coolant, and the following conditions exist:
Pressurizer level 99% and rising.
RCS pressure 1400 psig and rising.
PRT level is 73% and pressure is 5 psig and stable.
TI-453, PORV downstream temperature, is 138°F and rising.
TI-455, Safety Valve downstream temperature, is 125°F and rising.
TI-457, Safety Valve downstream temperature, is 125°F and rising.
TI-459, Safety Valve downstream temperature, is 126°F and rising.
Containment Pressure 0.96 psig and rising.
Containment sump level is rising slowly.
R-2, 7,11 and 12 are in alarm. Which one of the following states only potential sources of the RCS leak indicated by the given conditions?

A. PORV leakby Safety valve leakby B. PORV leakby PRZR Level upper tap break C. PRZR Steam Space sample line break Safety valve leakby PRZR Steam Space sample line break PRZR Level upper tap break Page: 29 cj 200 1211412009 A -Incorrect. Both are incorrect, since the PRT parameters are unchanged after the event has been in progress for 15 minutes. If the PORVs or the Safeties had leaked by, the PRT parameters would be higher than they initially would. Plausible, since the downstream temperatures are higher than they were, but only slightly due to elevated ctmt ambient temp in the vicinity of the steam space break. If either of the PORVs or Safeties were leaking by, the tailpiece temperatures would be much higher than this. B -Incorrect. The first part is incorrect (see A). The second part is correct (see D). C -Incorrect. The first part is correct (see D). The second part is incorrect (see A). o -Correct. These are both correct, since per the indications (przr level high and pressure low and rising due to going solid on SI flow) there is a steam space break. This choice has parts which are similar to incorrect choices which would be Przr Liquid Space sample and Przr Level lower tap, steam space sample and upper tap are both steam space penetrations. Ran a 400 gpm steam space break from 100% power (IC-73) on the simulator laptop to validate these numbers. Drawing PID: 0-175037 SH 2 Previous NRC exam history if any: 00BAA2.26 OOB Pressurizer Vapor Space Accident AA2. Ability to determine and interpret the following as they apply to the Pres;urizer Vapor SpaceAccident: (CFR: 43.5/45.13) AA2.26 Probctlle PZR stean spa:e le9<:aJe paths other than PORV or code safety .... 3.1 3.4 Match justification: This question presents a scenario with symptoms given of a steam space break. The indications have similarities to either a PORV or code safety leaking by OR another leakage path other than the PORV or code safeties, and some differences. In this case the applicant must correctly identify the potential sources of a a steam space break which for these symptoms must be other than a PORV or a code safety leaking by. Objective: 1 LABEL AND ILLUSTRATE the Pressurizer System flow paths to includethe components found on Figure 3, Pressurizer and Pressurizer Relief Tank (OPS-40301 E05). Page: 30 of 200 1211412009 Date: 10/7/2009 I I I IQI I @BI --463 ----j R ... , R I I L_fLL_J \SI Q1B31V027B 3-GM88FNY

- 6" X 3" RE I o (.) (.) "" QV032 3/4-A78 1-8053 ---l I Time: 01 :37:54 PM 8010tlA !

80lOAB, I TE TE TC QV071 ORcl to I 80lOAD 80l0M 'II -------I 2" HCB-62 : 6-RC-2501R L 1" I _ _ SIS RHR PUMP # 80l0AH 1MB ** QV031A Tn 80l0AC 1 2-RC-2501 R 4-RC*2501 R--, I I I I S NC 2 I I Title: C:\Reference Disk\Exam Reference Disk\Drawings\D175037 -0002.cal Date: 10/7/2009 NO. 1C UPPER 3/8" X .065" s.s. TUBING TYPICAL FOR HV3179C, HV3180C, HV3181C Time: 01 :38:43 PM 3/8" X .065" s.s. TUBING ---,--t--t-t-l )0-17500 STEAM G FROM RE SEE MATCf-REV. NO. __ DATE ___ REV. NO. __ DATE ___ REV. NO. __ DATE ___ REV. NO. __ DATE ___ REV. NO. __ DATE __ _ REV. NO. 1 2 3 4 5 Title: C:\Reference Disk\Exam Reference Disk\Drawings\D17S009-0002.cal

12. 008K3.01 001/FNP BANK/RO/C/A 3.4/3.5/N/N/3/CVRIY Unit 1 is at 100% power and the following occurred:

TK-144, L TON HX OUTLET TEMP controller, demand failed high. Which one of the following describes the impact on the Letdown System Temperature, and the required action? A.. Higher Letdown temperature.
Isolate Letdown and place Excess Letdown in service. By* Higher Letdown temperature.
Place TK-144 in manual and adjust flow. C.* Lower Letdown temperature.
Isolate Letdown and place Excess Letdown in service. D.* Lower Letdown temperature.
Place TK-144 in manual and adjust flow. A -Incorrect.

The first part is correct (see B). The second part is incorrect (see B). Plausible, since "IF letdown flow cannot be reduced [in manual control], THEN this would be correct, but it is not the first strategy prior to manual control of temperature. B -Correct. This controller demand goes up to raise temperature, which at 100% demand, sends a full closed signal to the CCW to the Letdown HX valve. The first alarm to come in would be: ARP-1.4, DF1, L TON TO DEMIN DIVERTED TEMP HI [at 135°F] , and that ARP states to: "Take manual control of L TON HX Outlet Temp TK-144 and attempt to increase CCW flow to the Letdown Heat Exchanger." The OPS "Skill of the craft" policy also states that this is appropriate prior to attempting anything else. The ARP also states: "6. Adjust charging or letdown flow as required to reduce the letdown flow temperature. AND: 5. IF letdown temperature can NOT be reduced, THEN close LTDN ORIF ISO 45 (60) GPM Q1 E21 HV8149A, B, and C." C -Incorrect. This is incorrect since demand failing high causes the CCW to the Letdown HX valve to close (to raise Letdown Temperature). Plausible, since many valves open when demand goes to 100%. Also, if the valve did go open, it would cause boron absorption in the Mixed Bed demineralizer due to the cooler Letdown temperature. The second part is incorrect, but plausible. If the TK-144 valve could not be controlled in Manual, this would correct. o -Incorrect. The first part is incorrect (see C). The second part is correct. Per OPS "skill of the craft" policy, placing a controller in MAN from AUTO when necessary to Page: 31 d 200 1211412009 control parameters is always appropriate. One skill of the craft item, which may be performed as necessary without procedure guidance, is: "?? Adjusting pots and controllers, including transfer between AUTO and MANUAL, to maintain parameters within log spec or procedural specs." The other procedure guidance for a Low Letdown temperature is the requirement to maintain Reactor power <100% at all times. Ran on Simulator Laptop (IC-73) AT 100%. DF1 was the first alarm to come in (less than 30 secs). ARP-1.4, VERSION 48, DF1, LTDN TO DEMIN DIVERTED TEMP HI 3. Take manual control of L TON HX Outlet Temp TK-144 and attempt to increase CCW flow to the Letdown Heat Exchanger.

4. Adjust charging or letdown flow as required to reduce the letdown flow temperature.

5. I F cause for the eI evated temperature has been corrected, TH EN refer to FNP-1-SOP-2.1, CHEMICAL AND VOLUME CONTROL SYSTEM PLANT STARTUP AND OPERATION to return TCV143 to DEMIN. 6. IF letdown temperature can NOT be reduced, THEN close L TON ORIF ISO 45 (60) GPM Q1 E21 HV8149A, B, and C. NOTE: T r ansi ents that will require boration or dilution should be avoided if letdONn has been secured. 7. IF a ramp is in progress, THEN place turbine load on HOLD 8. Go to FNP-1-AOP-16.0, CVCS MALFUNCTION to address the loss of letdown flow. ARP-1.4, VERSI ON 48, DF5, VeT T EM PHI, 4. Adjust charging or letdown flow as required to reduce the Letdown Flow Temperature.

5. Adjust L TON HX Outlet Temperature

< 111°F. DRAWING D175039SH 2 Page: 32 eX 200 1211412009 Previous NRC exam history if any: 008K3.01 008 Component Cool i ng Wo1er System K3 Knowledge of the effect that a lcesar malfunction of the CCWS will have on the foilONing: K3.01 Loa:Is cooled by CCWS ........................................... 3.4 3.5 Match justification: This question presents a specific type of malfunction of the CCW system (Failure of the CCW to the Letdown HX control valve controller). To answer this question correctly, knowledge of the effect of this malfunction of the CCW system on the Load (Letdown) cooled by CCW is required. The effect is that LETDOWN temperature goes up due to the controller failure causing the CCW valve to the load (letdown) to go closed. The second part of the question and answers were added to gain 3 plausible but incorrect distractors. Objective:

7. DEFI NE AND EVALUATE the opero1ional implico1ions of normal / abnormal plant or equipment conditions associo1ed with the safe opero1ion of the CCW System components and equipment, to include the following (OPS-40204A07):

Normal control methods
Abnormal and Emergency Control Methods
A utomo1i c actuo1i on i ncl udi ng setpoi nt (exampl e 81, Phase A, Phase B, High Radio1ion, LOSP)
Protective isolo1ions such as high flow, low pressure, low level including setpoint
Protective interlocks
Actions needed to mitigo1e the consequence of the abnormal ity Page: 33 of 200 1211412009

\ I Date: 10/7/2009 Time 01 :47:56 PM 2 I 3 I 4 I 6 D-175040 (B-IO) BTRS REHEAT HX 113E708 r .. , LOC 1-3 3/4' V 1UQ-771A QV503B QV479A QV503A ...--1M41HM4I3/4* D QV479B QVI91A 1-8397A 3/4-X92D 3/4-CS-15IR "l-I Ifl '" -U I I '" I U

I (') (') 3/4'-HCD-241 QVI90 1-8399 3-X92D QV507B 3/4' D QV505A HEAT EXCHANGER ITEM, AHNR SAFETY TUBES 2A SHELLS 2B QIE21H004 @_F1FOI34A 150 150 REF. G QVI50A 1-8407A 3/4-T78 l-FT QV 1-6 3/4 1-"' __

V QV505B QV506A 1----.... --3/4' D QV506B , R II-;D/I 150 HI __ l-FY / ...... __ :: R [tJ: §:iiI50 ' 150 R 3/4' v QV467l / 3'-HCB-4 3-CS-15IR 2A D-175009 SH.1 (H-IO) SS 1-8390 3'-HCB-76 3-CS-15IR SEAL HEAT EXCHANGER

ITEM, SAFETY CLASS' TUBES-2A SHELL-2B QIE21H003 QVI95 1-8400 3-X92D 3/4' D 3/4-T78 3'-HCB-22 3 CS 51R QV194B 1-8398B 3-X92D QV468 1--.... 3/4* D QVl96 1-8482 3-X92D LC 3/4' D QV477 3'-HCB-74 f1 QV262 3-CS-151R 1-8123 3/4' VENT QV469 3'-HCB-25 QV261

1-8120 :3-RV72LNS
- - - - 1 3/4' D :1 CVCS EVAP FEED PUMP 110E386 SH.4 LOC. (E-8) D-175039 SH.4 (E -6) QV265 l_Q1n,::.

3'-HCB-63 QVI97 1-8484 3-X92D enl'" --Ilfl u l T'" en ., '" U I I Iv, 0 ., C :i '" C\J " C> Ifl I '" A >-'Ifl 2:::cru "-"'I <[ ",,,, ",,,,. (')U ",W D a..O...J ;.::! QV029 X I WWU' !l.._---1 1 , Title: C:\Reference Disk\Exam Reference Disk\Drawings\D175039-0002,cal 04/03/09 13 :21 :40 FNP-I-ARP-l.4 LOCATION DFI SETPOINT: 135 D F ORIGIN: I-TY-143X Auxiliary Relay actuated by Temperature Bistable (NIE21 TB143) LTDNTO DEMIN TEMP HI PROBABLE CAUSE 1. Low or Loss of CCW Flow to the Letdown Heat Exchanger.

2. Letdown Flow greater than Charging Flow. AUTOMATIC ACTION 1. Letdown High Temperature Divert Valve QIE21 TCV143 diverts Letdown Flow to the V CT. {CMT 0008644 } OPERA TOR ACTION 1. Verify QIE21 TCV143 has diverted letdown flow to VCT to bypass demins 2. Monitor charging and letdown flows and temperatures.

4. Take manual control ofLTDN HX Outlet Temp TK-144 and increase CCW flow to the Letdown Heat Exchanger.

Adjust charging or letdown flow as required to reduce the letdown flow temperature. PTn,,,,," temperature can NOT be reduced, THEN close L TDN ORIF ISO 45 (60) GPM QIE21HV8149A, B, and C. NOTE: Transients that will require boration or dilution should be avoided if letdown has been secured. 6. IF a ramp is in progress, THEN place turbine load on HOLD 7. Go to FNP-I-AOP-16.0, CVCS MALFUNCTION to address the loss of letdown flow.

References:

A-177100, Sh. 206; D-175039, Sh.2; D-I77091; D-I77375; U-175997; PLS Document Page 1 of 1 Version 45.0

13. 008K4.09 001/MOD/RO/C/A

-2.7/2.91IN/2JCVRIY Unit 1 was operating at 100% power, and the following conditions occurred: At 1000:

A Train is the "On Service" train.
1 B CCW pump is running and supplying loads in the on-service train.
1A CCW pump is running to support charging pump operations.
1 C CCW pump is aligned and OPERABLE.

At 1005:

A Safety Injection and LOSP occurred simultaneously.

Which one of the following combinations of CCW pumps will be running following the operation of the ESF sequencers, with no operator actions? A'! 1A and 1 C CCW pumps ONLY. B. 1 Band 1 C CCW pumps ONLY. C. 1A and 1 B CCW pumps ONLY. D. 1A and 1B and 1C CCW pumps. A -Correct. 1 B CCW pump is running on A train but will trip on Load shed. Then, the auto start circuitry starts up the non-swing, train related 1A & 1C pumps per CCW FSD Appendix A step 3.1.2.2, LOSP. B -Incorrect. 1 B CCW pump is running on A train and if there was no LOSP signal, the SI auto start circuitry would leave the 1 B running and not start the pump on the same train. The opposite train pump is 1A, and not 1C. This is plausible since this is the opposite train and CCW has backward logic. normally 1 A pump would be assigned to A train, but CCW is an exception to this general rule. C -Incorrect. Plausible, since this would be correct with an SI and no LOSP. 0-Incorrect. Plausible, since the SW pumps would have all pumps including the swing running in the event of an SI if the swing pump was running to start with. For this CCW system alignment: 1 B CCW pump is running on A train and 1A CCW pump is running on B train to start with. For an SI alone, 1 Band 1A would be left running. For an LOSP, 1A and 1 C would be started. However, the LOSP sequencer secures 1 B prior to starting 1 C. FNP-1-SOP-23.0, Version 83.0 3.2 CCW is normally lined up so that

One CCW pump and one CCW heat exchanger is in operation supplying the on-service train and the secondary heat exchangers.
The remaining pump and heat exchanger are valved into a closed loop with the redundant safety train. The off-service trai n is normally in operation in modes 1-4 supplying the operati ng chargi ng pump, wi th the non-operati ng SFP HX flowpath aI i gned and CCW to the Page: 34 of 200 1211412009 RHR HX isolated. (Reference RER 1080944901)

FSD A-181 000 Appendix A 3.1.1.3 SIAS In the event of a SIAS with offsite power available, the on-service pump shall continue to operate, and the off-service (redundant) train-dedicated pump shall automatically start. Swing pump B shall continue to provide backup in the event of a fault trip of the dedicated pump in the train to which swing pump B is aligned, as above (Reference 6.7.11). 3.1.2 Swing Pump B on-Service, Dedicated Pump Available (Possible Alternate Alignment To Equalize Pump Wear) 3.1.2.1 On-Service Pump Trips a. During normal plant operation, if on-service pump B trips due to a fault, the dedicated pump in the on-service (operational) train shall automatically start and supply component cooling water to the on-service component cooling heat exchanger (Reference 6.7.11). 3.1.2.2 LOSP In the event of a LOSP with or without a SIAS, on-service pump B shall be shed and the two train dedicated pumps, C and A, shall be automatically sequenced onto the diesel generators (Reference 6.7.11). Swing pump B shall provide backup in the event of a fault trip of the dedicated pump in the train to which swing pump B is aligned (Reference 6.7.11). 3.1.2.3 SIAS In the event of a SIAS with offsite power available, on-service pump B shall continue to operate and the off-service train-dedicated pump shall automatically start. The dedicated pump in the on-service (operational) train shall continue to provide backup in the event of a fault trip of swing pump B (Reference 6.7.11). Page: 35 of 200 1211412009 Previous NRC exam history if any: 008K4.09 008 Component Cool i ng Water System K4 Knowledge of CCWS design feature(s) and/or interlock(s) which provide for the follONing: (CFR: 41.7) K4.09 The "staldby" feEturefor theCCW pumps ............................ 2.72.9 Match justification: Objective:

2. RELATE AND I DENTI FY the operational characteristics including design features, capacities and protective interlocks for the components associated with the CCW System, to include the components found on Figure 2, Component Cooling Water System, Figure 3, Secondary Heat Exchanger Header, and Figure 5, RCP-CCW & 8W System (OPS-40204A02) . 7. DEFINE AND EVALUATE the operational implications of normal / abnormal plant or equipment conditions associated with the safe operation of the CCW System components and equi pment, to i ncl ude the foil owi ng (OPS-40204A 07):

Normal control methods
Abnormal and Emergency Control Methods
A utomati c actuati on i ncl udi ng setpoi nt (exampl e 81, Phase A, Phase B, High Radiation, LOSP)
Protective isolations such as high flow, low pressure, low level including setpoint
Protective interlocks
A cti ons needed to mi ti gate the consequence of the abnormal i ty Page: 36 of 200 1211412009 FNP UNITS 1 & 2 CCW SYSTEM A-181000 1.0 2.0 3.0 3.1 3.1.1 3.1.1.1 APPENDIX A CCW PUMP ALIGNMENTS PURPOSE The following provides the functional requirements for the various CCW pump alignments.

There are no new requirements in this Appendix in addition to those in the FSD text. The purpose of this Appendix is to extract the requirements from the FSD which pertain to pump alignments. INTRODUCTION CCW pumps C and A shall be train dedicated and aligned to the 4 kV buses F (train A) and G (train B) , respectively. CCW motor pump B, when available, shall be aligned to either of the vital 4 kV buses, F (train A) or G (Train B), corresponding to whichever train the B pump has been valved into. The B pump shall be physically valved into the train from which it is set to receive its power. CCW PUMP ALIGNMENTS ALL THREE CCW PUMPS OPERABLE One Train-Dedicated Pump On-Service, Swinq PumP B in Standby (Aligned to the On-Service Train), One Dedicated Pump Off-Service (Nor.mal System Alignment) On-Service Pump Trips a. During normal plant operation, if the on service pump trips due to a fault, the standby pump shall automatically start and supply component cooling water to the on-service component cooling heat exchanger (Reference 6.7.11). b Once the (swing) pump B becomes the on-service pump, the breaker of the pump with the fault is required to be racked out immediately, or the lock-out relay shall not be reset by the operator, in order to allow the proper operation of swing pump B in the event of an LOSP (Reference 6.7.11) c Deleted (Reference 6.7.057) 3.1.1.2 7597\A-181000.SD A-1 Rev. 5 FNP UNITS 1 & 2 CCW SYSTEM A-181000 3.1.1.3 3.1. 2 3.1.2.1 3.1.2.2 of the dedicated pump in the train to which swing pump B is aligned (Reference 6.7.11). SlAS In the event of a SIAS with offsite power available, the on-service pump shall continue to operate, and the off-service (redundant) train-dedicated pump shall automatically start. Swing pump B shall continue to provide backup in the event of a fault trip of the dedicated pump in the train to which swing pump B is aligned, as above (Reference 6.7.11). Swing Pump B on-Service, Dedicated Pump Available (Possible Alternate Alignment To Egualize Pump Wear) On-Service Pump Trips a. During operation, if on-service pump B trips a faulh tjie __ dedicated on-serVlce (operational) train shall --}UEomatically staEt and __ water to the on-serVlce component I2:Kchanger (Reference 6.7.11) --b. Deleted (Reference 6.7.057) LOSP ---------3.1.2.3 In the event of a LOSP with or without a SIAS, -9n-service 12ump B be shed and two dedicated pumps, C and A, shall be automatically sequenced onto the diesel enerators 6.7.11) . Swing pump B shall provide backup in the event of a fault trip of the dedicated pump in the train to which swing pump B is aligned (Reference 6.7.11). SlAS (lIn the event of a SIAS with offsite power available, ..fo£A on-service urn B shall continue to .2J2erate_and the " \: 0 -service train-deaicated pum12 shall . tare. Tne(credlcated pump in the 7597\A-181000,SD train shall continue to provide backup in the event of a fault trip of swing pump B (Reference 6.7.11). t m (12 \ A-2 Rev. 5 FNP Units 1 & 2 3.1.4.4 3.1.4.5 SERVICE WATER SYSTEM A-181001 are required between pumps 1 Band 1 C and between pumps 1 C and 1 D to separate trains for fire protection purposes. (Reference 6.1.010) Each pair of train oriented Service Water pumps along with the swing SW pump shall be provided with a minimum flow bypass valve (See Section 3.4 for the required flow rates) to recirculate service water to the service water intake structure wet pit. (References 6.4.014 and 6.4.018) Each Service Water pump motor shall be equipped with bearing temperature monitoring devices. (Reference 6.5.003) 3.1.5 I & C Requirements 3.1.5.1 The Service Water pumps shall be automatically started by a signal from the LOSP or ESS sequencer. The Service Water swing pump shall be automatically "'"Slgi1aIfrom the sequencer-when in service _ .-----replacing one of the train oriented pumps. (References -t-_**-L ;f J -.Y1,\ v\ f-ce -6.7.039 and 6.1.00J3)

L) t: (, k' .

{e _.'TT-} -,' v.1 ) I er kJ2. ).C c/(a,/ Q<J-e .* V' !.J-4// c 3.1.5.2 Key interlocking of power supply breakers, disconnect 3.1.5.3 3.1.5.4 3.1.5.5 switches, and SW header cross-connect valves shall be f c.t. A f\ used to ensure alignment of the Service Water swing pump to one train only. (Reference 6.1.006) Annunciation shall be provided in the Control Room to alert the operator when a Service Water pump breaker trips. (References 6.4.1 04 through 6.4.115, 6.1.007) Monitor lights shall be provided in the control room to allow quick verification of the status of Service Water Pumps A, B, D, and E following a safety injection signal. (Reference 6.7.124) The Service Water (SW) System shall have redundant level instrumentation to monitor and control the Storage Pond/Service Water Pump Wet Pit level. (Reference 6.4.075) Service SW Pump Wet Pit SW Pump Wet Pit 3-5 A B TPNS Nos. Nl(2)P25LI4066A Nl(2)P25LI4066B Rev.44 I

1. CCW-40204A07 013IHLT/IMEM 2.712.9/00SK4.09////00SA3.0S 008K4.09 Unit 1 is operating at 100% power with the following conditions:

"A" Train is the "On Service" train.
1 B CCW pump is running and supplying loads in the on-service train.
1A CCW pump is running to support charging pump operations.
1 C CCW pump is aligned and OPERABLE.

A Safety Injection occurs at this time. Which one of the following combinations of CCW pumps will be running following the operation of the ESF sequencers? (Assume no operator action is taken) A. 1A and 1 C CCW pumps ONLY B. 1 Band 1 C CCW pumps ONLY C'!' 1A and 1 B CCW pumps ONLY D. 1A and 1B and 1C CCW pumps Page: 1 of3 10/26/2009 008K4.09 008 Component Cooling Water System (CCWS) K4 Knowledge of CCWS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) K4.09 The "standby" feature for the CCW pumps ....... 2.72.9 A. Incorrect, 1 B CCW pump is running on A train therefore, 1 C CCW pump will not start. This is plausible since on a load shed and LOSP sequencer operation this would occur. B. Incorrect, 1 B CCW pump is running on A train therefore, 1 C CCW pump will not start, 1A pump will start on B train. This is plausible since this is the opposite train and CCW has backward logic. C. Correct, 1 B CCW pump is running on A train therefore, 1 C CCW pump will not start. 1A CCW pump is running and does not recieve trip signal and it remains running. D. Incorrect, 1 B CCW pump is running on A train and 1A CCW pump is running and would recieve a start signal from B train sequencer. The 1 C CCW pump will not start since the 1 B CCW pump is running. This is plausible since the SW pumps would have all pumps running in this situation. Lesson plan ops-52102G The three CCW pumps (Figures 6 & 7) can be operated from the MCB or locally at the HSP by a three-position handswitch (STOP/AUTO/START, spring return to AUTO). A two-position selector switch (LOCAL/REMOTE) at the HSPs determines which station has control of the pumps. The dedicated pumps (A and C) will automatically start on receiving an "S"-signal or a loss of offsite power (LOSP) signal, provided that the local remote selector switch is in the REMOTE position and the MCB handswitch is in the AUTO position. The swing pump (B) acts as a backup for the dedicated pumps by being mechanically and electrically aligned to the same train as one or the other of the dedicated pumps. The swing pump will automatically start when: (1) the dedicated pump it is backing up trips on overload; (2) the selector switch is in the REMOTE position; (3) the MCB hand switch is in the AUTO position. The swing pump also receives start signals from the safety injection (SI) sequencer and the LOSP sequencer. However, it will only start if the selector switch is in REMOTE, the MCB switch is in AUTO, the dedicated pump A or C (depending on which train it is lined up to) has tripped on overload, or its supply breaker has been racked out. FSD A-181000 3.1.5.2 During normal plant operation, with all pumps operational, if the operating pump power supply breaker trips, the standby pump shall automatically start and supply CCW to the CCW heat exchanger in operation. The breaker of the pump with the fault shall be racked out immediately, or the lockout relay shall not be reset by the operator in order to allow the proper operation of the standby pump in the event of a loss of offsite power (LOSP). The CCW pump overload trip shall be alarmed in the MCR to alert the operator (References 6.1.01,6.4.15,6.4.16,6.4.17). Page: 20f3 10/26/2009 2008 NRC exam Technical

Reference:

FSD A-181000 Learning Objective: 40204A07 List the automatic actions associated with the Component Cooling Water System components and equipment during normal and abnormal operations including (OPS40204A07):

Normal control methods
Automatic actuation including setpoint (example SI, Phase A, Phase B, High Radiation, LOSP)
Protective isolations such as high flow, low pressure, low level including setpoint Protective interlocks also 52102G02 Comments:

This meets the KA since it tests the standby feature of the standby pump for the train it is aligned to. This is the standby feature for the main pump (ie., 1C or 1A CCW pump). Our SW pumps do not have a feature where the standby pump looks to see if the other pump is running before starting or not starting the other pump in that train for an SI signal. In that case there would be 5 SW pumps running. For CCW, if the swing pump is running, then the other pump in that train will not start on the SI signal. All distracters are plausible since our trains are not set up in a logical way and C CCW pump is A train and A CCW pump is B Train. Most other components are configured correctly and differently. Had to change the stem to take into account the new CCW and charging pump line up. FNP BANK: CCW-52102G02 05 Page: 3 of3 10/2612009

14. 009EG2.1.23 001/NEW/RO/C/A 4.3/4.4/N/N/4/CVRNER 5 EDITORIAL Unit 1 has experienced a Small Break LOCA, and the following conditions occurred:

At 1000:

ESP-1.2, Post LOCA Cooldown and Depressurization, is in progress.
Normal Charging has been established.

At 1010:

CTMT Pressure is 6 psig and rising.
Subcooling is 24°F and decreasing.
PRZR Level is 28% and decreasing.

Which one of the following is the required action lAW ESP-1.2? A. FK-122 must be adjusted to raise Przr level. B. Place the SI ACTUATION switch to ACTUATE. C. FK-122 must be adjusted to maintain current Przr level. Dy HHSI flow must be established and additional CHG PUMPs started. A -Incorrect. The Fold Out Page requires reinitiating HHSI flow due to both Subcooling and PRZR Level being too low with adverse numbers "16°F{45°F} & 13%{43%}. Plausible, since this would be correct if the procedure step for maintaining pressurizer level was initiated with adverse numbers and 50% PRZR level was required, while forgetting about the FOP requirement to re-establish HHSI flow. B -Incorrect. (see D). Plausible, since HHSI flow is needed, and it may seem more convenient to turn the SI switch vice going to the attachment to manipulate each component, but the FOP requires the attachment be used. This ensures that only the SI equipment and Phase A components desired are manipulated. C -Incorrect. (see A). Plausible, since if Adverse numbers were not taken into account, this would be correct per step 20.2.1 of ESP-1.2. D -Correct. The FOP requires this for these Subcooling and Przr level values. RO knowledge requires knowing the FOP requirements. ESP-1.2, Revision 23 Page: 37 of 200 1211412009 Previous NRC exam history if any: 009EG2.1.23 009 Small Break LOCA 2.1.23 Ability to perform specific system and integrated plant procedures during all modes c:I plant operation. (CFR: 41.10/43.5/45.2/45.6) RO 4.3 SRO 4.4 Match justification: This question requires knowledge of specific system and integrated plant procedures (ESP-1.2 Fold out Page) during a SBLOCA to answer correctly. Objective:

6. EVALUATE plant conditions and DETERM I NE if any system components need to be operated while performing ESP-1.2, Post LOCA Cooldown and Depressurization. (OPS-52531 F06) Page: 38 of 200 1211412009 FNP-I-ESP-l.2 POST LOCA COOLDOWN AND DEPRESSURIZATION Revision 23 Step n NOTE: 7 20.2 Action/Expected Response Response NOT Obtained During a LOCA, a full or rlslng pressurizer level may indicate a steam space LOCA exists. In that event, step 20.2, RNO provides gUidance if the RCS must be operated water solid, and charging used to maintain subcooling instead of pressurizer level. ""':'--------[CAl Maintain level greater than 25%{50%}.

20.2 IF solid plant operation required, THEN perform the following. 20.2.1 IF charging flow path aligned, a) Maintain SUBCOOLED MARGIN THEN control charging MONITOR indication greater than 26°F{55°F}. CHG FLOW FK 122 b) Control charging flow to stabilize subcooling at , existing value. 1\'\ CHG FLOW {C [ 1 FK 122 adjusted ************************************************************************************** CAUTION: To prevent potential seal damage, neither seal injection nor CCW cooling should be restored to a RCP which has lost both seal injection and CCW cooling. ************************************************************************************** _21 [CAl Check if RCP(s) should be reconfigured to optimize RCS flow and pressurizer spray performance. 21.1 Check RCP IB -STOPPED. 21.1 Perform the following. 21.1.1 Verify RCPs lA AND lC -STOPPED. RCP [l lA [l lC 21.1.2 Proceed to step 22. OBSERVE CAUTION AND NOTE PRIOR TO STEP 22. Step 21 continued on next page. ___ Page Completed Page 25 of 50 r"""? ....--------.-----------'"""'rh.:t,,:)-i:rir,,- -.,7----------.,.-----------.

eAGE FNP-1-ESP-1.2 POST LOCA COOLDOWN AND DEPRESSURIZATION Revision 23 Step n 1 2 1.1 Action/Expected Response Response NOT Obtained I I I __ t /'_ I L;;,C'(C-e <-Monitor SI reinitiation criteria following HHSI isolation. Greater than 1""..J./

subcooled in PRZR additional CHG PUMPs as level above 43%1) required using ATTACHMENT

5. ' PRE-ESTABLISHING HHSI FLOW. \ I ) Monitor FNP-I-EEP-2 and FNP-I-EEP-3 branch criter1a.

2.1 No SG pressure falling in an uncontrolled manner or less than 50 psig. 2.1 IF affected SG NOT previously isolated. THEN go to FNP-1-EEP-2. 2.2 No high secondary radiation or SG level rising uncontrolled. 2.2 Establish HHSI flow. and start additional CHG PUMPs as required using ATTACHMENT

5. RE-ESTABLISHING HHSI FLOW THEN go to FNP-1-EEP-3.

3 Monitor switchover criteria. 3.1 RWST level greater than 12.5 ft. 3.2 CST level greater than 5.3 ft. 3.1 Go to FNP-1-ESP-1.3. 3.2 Align AFW pumps suction to SW using FNP-1-S0P-22.0. 4 Monitor charging miniflow criteria (during SIlo 4.1 RCS pressure less than 1900 psig. 4.2 RCS pressure greater than l300 psig. 5 Monitor adverse containment criteria. 5.1 CTMT pressure less than 4 psig and radiation less than 10 5 R/hr. 4.1 Verify miniflow valves open. 4.2 Verify miniflow valves closed. 5.1 Utilize bracketed adverse CTMT condition numbers.

15. 010K1.03 001/NEW/RO/C/A

-3.6/3.7/N/N/4/CVRN Unit 1 was at 28% power and the following conditions occurred:

All PRZR Backup Heaters are in AUTO.
A CVCS Malfunction has occurred.
FK-122, CHG FLOW, has been placed in manual.
PRZR level is at 36% and rising. Which one of the following describes the operation of the Backup Heaters and the Spray valves with no operator actions? All PRZR Backup Heaters will be (1) and BOTH PRZR Spray Valves will be _---1.!:(2:..t..)

__ (1) Backup Heaters (2) Spray Valves A'! ON Opening B. ON Closing C. OFF Opening D. OFF Closing Page: 39 of 200 1211412009 A -Correct. First part: The CVCS Malfunction caused an insurge, which caused the PRZR level in increase. PRZR level program is 21.4-50.2% level from 547-573°F Tavg, so at 28% power, program level is 29.5% przr level. There is a 6.5% level deviation (>5%). Przr level >5% above the program level turns on all BU heaters which cause the pressure to go up more (after the water reaches the new higher saturation temperature). Second part: The insurge caused the PRZR steam space to be compressed, which causes the Pressure go up. The pressure controller opens both spray valves until pressure stabilizes. While pressurizer level is increasing and all backup heaters are on, pressure will continue to increase and spray valves will continue to open. B -Incorrect. First part correct (see A). Second part incorrect. Plausible, since subcooled water has insurged into the pressurizer, and a cooler steam space would cause pressure to decrease, but the compression of the steam space in the pressurizer due to the increasing level raises pressure and overrides the cooler temperature of the pressurizer liquid which would tend to lower pressure. C -Incorrect. First part incorrect. Plausible, due to the pressure going up. This automatically turns off all Backup heaters unless there is a > 5% high level deviation as in this case. Second part is correct (see A). D -Incorrect. First and second parts incorrect. Plausible, since an error in the second part (thinking that the subcooled water would drop pressure in the pressurizer) combined with a miscalculation of program level, or using the 100% value of program level, would indicate a PRZR level less than program and pressure low. These errors would cause this choice to be selected. Second part is incorrect (see B). ARP-1.8, Version 33.0 Drawing 0175037 Sheet 2 Ran this malfunction on the simulator laptop: IC-38, 27% Power, when PRZR level increased to 6% above program level all Backup heaters were on and spray valve demand had increased from the initial value of 6.6% to 20% open. The spray valves continued to open further for several more minutes. Page: 40 of 200 1211412009 Previous NRC exam history if any: 010K1.03 010 Pressurizer Pressure Control System K 1 KnOllVleclge of the physical connectionsandlor cause-effect relationships between the PZR pes and the foilOlNing systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8) K1.03 RCS ........................................................... 3.63.7 Match justification: A CVCS malfuction in this question causes excess mass in the RCS which causes an insurge into the pressurizer. This causes a level deviation in the pressurizer which energizes the pressurizer heaters even though pressure is high, and opens spray valves due to the steam space compression and rising pressure, even though the insurge water is subcooled. Objective:

11. Given a set of plant conditions, LIST AND DESCRI BE theactions'effects that will occur following a CVCS Malfunction with no operator action (OPS-52201H15).

Page: 41 of 200 1211412009 11125/087:43:37 FNP-I-ARP-I.8 LOCATION HA2 .. " .. ORIGIN: Level Bistable LB-459D from Level Transmitter LT-459 orLT-461 and TY-408 median TAVG. PROBABLE CAUSE PRZRLVL DEVHI BIUHTRS ON 1. Pressurizer Level Instrument or Control System malfunction.

2. Plant Transient while in manual rod control. 3. Rod Control System malfunction.

4. Charging or Letdown System malfunction.

Pressurizer Backup Heaters energize. OPERATOR ACTION 1. Place turbine load on HOLD. 2. Check pressurizer level indications and determine the actual level deviation.

3. IF an instrument failure has occurred, THEN go to FNP-I-AOP-I00, INSTRUMENT MALFUNCTION.

4. Ensure that the pressurizer backup heaters are energized.

5. IF required, THEN take manual control ofCHG FLOW FK 122 and decrease charging flow to return pressurizer level to the program band. 6. Determine the cause of the level deviation by checking:

5.1 Charging flow 5.2 Letdown flow 5.3 BTRS flow 5.4 Charging pump status 7. IF the alarm was caused by a plant transient, THEN control the transient and return Pressurizer Level to normal. 8. IF a charging OR letdown system malfunction exists, THEN go to FNP-I-AOP-16.0, CVCS MALFUNCTION.

References:

A-I77100, Sh. 357; U-260610; D-177109; D-177111; D-177112; D-177113; U-26664 7 PLS Document; Technical Specifications Page 1 of 1 Version 32.0

16. 011 K 5.05 001/NEW/RO/C/A 2.8/3.1/N/N/2ICVRIY Unit 1 is at 100% power, and has experienced a Pressurizer Level Control Malfunction due to the controlling pressurizer level transmitter failing. The following conditions exist:

PRZR LVL CONT CH, LS/459Z, is in the "1/11 LT459/60" position.
Tavg is 573.0°F.
AOP-100, Instrumentation Malfunction, is in progress.
Przr level is 40% and rising.
Przr level control is in Manual.
Charging flow is 125 gpm.
Letdown flow is 130 gpm.
Seal Injection flows are:
Seal Leakoff Flows are: Which one of the following is the: 1A 8.1 gpm 2.9 gpm 1B 7.9 gpm 3.0 gpm 1C 8.0 gpm 3.1 gpm. 1) approximate time that it will take for the Pressurizer level to get to program level at the current rate in Manual control, and 2) the correct switch position for PRZR LVL CONT CH LS/459Z lAW AOP-100? Switch position A. 56 Minutes lIlli, L T459/61 B. 94 Minutes lIlli, LT459/61 Cy 56 Minutes 111111, L T461/60 D. 94 Minutes 111/11, LT461/60 A -Incorrect.

The time is correct (see C). The second part is incorrect, since L T-459 was the controlling channel, and it needs to be selected completely out by selecting 111111, 111/11, L T461/60. This will place the remaining two operable L Ts in service for Pressurizer level control. Plausible, since confusion may exist as to which of the two selected channels controls pressurizer level and which performs other control functions in relationship to the switch position. For example, if L T-460 was the failure this would be correct. B -Incorrect. The time is incorrect, since the level of the pressurizer in percent is a volume affected by the specific volume at normal Operating Pressure Pressurizer Temperature (about 648°F). Plausible, since the pressurizer curve lists the change in level from 40-50% at 93 gals I %, but charging 100°F water of 56 gal volume will expand to 93 gallons for a 1 % rise. The second part is incorrect (see A). Page: 42 of 200 1211412009 C -Correct. The time is correct, with a 100% program przr level of 50.2%, since a properly performed flow balance calculation shows that there is 10 gpm more charging into the RCS (in Charging and Seal inj minus the seal leakoff) than is leaving (in Letdown). Due to the specific volume of water at charging system temperature (about 100°F), which expands to pressurizer temperature (about 650°F), 56.3 gallons of charging water will equal 1 % in the pressurizer (Per STP-9.0, RCS Leakrate determination). (1 min/10 gals) * (56.3 gals/%) * (10%) = 56.3 mins This gallons/% level relationship is also verified by steam table Specific Volume calculation (see below). D -Incorrect. The time is incorrect (see B). The second part is correct (see C). At 100°F, the specific volume of water is 0.016130 ft3/1bm per the steam tables, and it would expand to 0.02657 ft3/1bm @ pressurizer temperature of 648°F. The charging water of 56 gals/min will expand to 93.5 gals/min at PRZR temperature (which is 1 % in the Pressurizer per Tank Curve 42). In approximately 56 minutes, at 10 gpm net Charging flow into the RCS, the pressurizer level will rise 10%. Tank Curve: Unit 1 Volume II Curve 42 (Hot Calibrated) 40% level=4403.04 gals 50% level=5336 gals 93.5 gals/% PRZR level Hot calibrated (650°F) Per Steam Table: 0.016130 ft 3/1bm @1 OO°F 0.02249 ft 3/1bm @ 575°F 0.02657 ft 3/1bm @648°F (Charging water expands in the RCS, which causes a pressurizer insurge, which in turn expands further in the pressurizer). PRESSURIZER PRESSURE AND LEVEL CONTROL, OPS-62201 H, OPS-52201 H, ESP-52201 H, Student Text, Figure 8 Page: 43 d 200 1211412009 LEVEL % 21.4 / , 547 Previous NRC exam history if any: 011 K 5.05 011 Pressurizer Level Control System 573,0 MEDIAN T AVG K5 Knowledge of the operational implications of the foilaNing concepts as they apply to the PZR LeS: (CFR: 41.5/45.7) K5,05Interr*!ciion of indicaro chcrging flow rate with volume of water requirro to bring PZR leII*! bed< to progrcrnmro leII*! hot/cold ................. 2.83.1 Match justification: This question requires knowledge of determining what the net charging flow into the RCS is, and then determining the time for the pressurizer level return to program setpoint. The pressurizer level program value has been provided to ensure it is clear which program level is being used in this question. Program level changes each cycle and with changes in Tavg throughout each operating cycle, so it is provided. To obtain 3 plausible but incorrect distractors, a second part was added to test the system knowledge of the Pressurizer control system selector switch. Objective:

5. DEFI NE AND EVALUATE the operational implications of abnormal plant or equipment conditions associated with the operation of the Pressurizer Pressure and Level Control System components and equi pment to i ncl ude the foil owi ng (OPS-52201 H07):

Normal Control Methods
Abnormal and Emergency Control Methods
Automatic actuation including setpoint, if applicable
Protective Interlocks A cti ons needed to mi ti gate the consequence of the abnormal i ty Page: 44 of 200 1211412009 05/12/09 12:23:30 FNP-1-AOP-100 6 T 59 LI 459Z HSDP INSTRUMENTATION MALFUNCTION 462 LI 462 COLD-CALIBRATED LER LI ROOM 459B LK 459 MEDIAN T AVG LI 459A MCB CHARGING FLOW CONTROL FCV 122 SECTION 1.2 Figure 1 LR459 DEVIATION LO LVL ALARM DEVIATION HI LVL ALARM; B/U HTRS ON ISOLATE LCV 8149A,B,C; TURN OFF ALL HTRS; MCBALARM PRESSURIZER LEVEL PROTECTION AND CONTROL Page 1 of 1 Version 7.0 OpsPcs113 HI PRZR LVL RX TRIP PRZR HI LVL ALARM Approved 100 90 80 70 .... 60 c CD 50 u '-Q) 40 11. 30 20 10 0 663.18 Pressurizer (Q1 B31 K001) Capacity vs. % level Unit 1 Volume n Curve 42 Hot Calibrateq*

Date-#- %level@.allQns o 663.18 10 1598.15 20 2533.11 25 3000.59 30 3468J!8 4403.04 338.00 62n97 7207.94 7675.42 8142:90 9071.87 10012.83 0 2000 4000 6000 8000 Gallons "Based on saturated liQuid temperature at 2235 psig Calculation Ref. No. SJ-OO-2200..Q03 --5-""3-'? f, 17 C --/ 1000

/('2t" ___ 93. Sf; od,'" 1% PRZR PROGRAM LEVEL % 50.2 ---------------------------------------------------

21.4 547 573.0 Pressurizer Program Level Figure 8 OpsPrs013 MEDIAN T AVG OPS-6220 H/52:

17. 012K2.01 001/FNP BANK/RO/C/A 3.3/3.7/N/N/3/CVRIY A loss of A Train Auxiliary Building 125V DC Bus has occurred on Unit 1. If the plant experienced a problem which required manually tripping the reactor, which one of the following describes the effect (on any closed Reactor Trip and/or Bypass breakers) of placing the RX TRIP ACTUATION switch on the MCB to TRIP? Placing the MCB handswitch in TRIP would ____ if they were closed. A'I open ALL reactor trip and bypass breakers.

B. ONLY open the 'A' reactor trip breaker and the 'B' reactor trip bypass breaker. C. ONLY open the 'B' reactor trip breaker and the 'A' reactor trip bypass breaker. D. open BOTH reactor trip breakers but NOT open either reactor trip bypass breaker. A -Correct. Aux Building DC power is not required to trip open breakers, as long as the UV coils are deenergized by Solid State (SSPS). Voltage from SSPS feeds the 48V UV coils that will allow the trip breakers to open when power is removed (a trip signal deenergizes the UV coils). Loss of "A" train AB DC would prevent the closure of the A RTB & B RTBYP breakers, AND would prevent the shunt trip coils on the A RT & B BYP breakers from being energized to provide an additional trip signal. SSPS power is from the inverters which supply power from the Regulated AC, bypassing the inverters, if AB DC is lost. B -Incorrect. See A. Plausible, since the Reactor Trip and Bypass breakers are operated and tripped by opposite trains. However, these two breakers are both operated by the B train aux building DC, and not the A train. Also, the Shunt trip coils operate to trip these breakers and the coils get power from AB DC (B train). However, the UV coils can still deenergize if needed and trip all of the reactor trip breakers. Confusion may exist as to which train of breaker is operated by which train of DC, AND as to which type of DC is needed to trip the breaker (UV coil 48V or Shunt Trip coil 125 V). C -Incorrect. See A. Plausible, since the Reactor Trip and Bypass breakers are operated and tripped by opposite trains, AND these two breakers are both operated by the A train aux building DC. Also, the Shunt trip coils operate to trip these breakers and the coils get power from A train AB DC. However, the UV coils can still deenergize if needed and trip all of the reactor trip breakers. Confusion may exist as to which type of DC is needed to trip the breaker (UV coil48V deenergizing or Shunt Trip coil 125 V energizing). D -Incorrect. See A. Plausible, since the AB DC does supply the shunt trip coils, and only the local pushbutton energizes the Shunt trip coil to trip the Bypass breakers, so there is a difference in the way the trip breakers and the bypass breakers work for loss of AB DC. However, the UV coil will still trip all RT & BYP breakers if a manual trip is called for. Confusion may exist as to the redundant methods using the UV and Shunt Trip coils to trip the reactor. Page: 45 c:J 200 1211412009 Reactor Protection Functional System Diagram (FSD) A181007, section 3.3.2 Eoch circuit breaker shall be equipped with a48 volt DC instantaneous undervoltaJetrip device and a 125 Vdc shunt trip device. (Reference 6.4.086) The Shunt Trip Attochment coil shall operate on 125 V dc and functi on as a backup for the undervoltaJe tri p devi ceo The first method of tripping the breaker (i.e., reactor trip or bypass breakers) is by a loss or drop of rated voltage to the Undervoltage Relay (UV). The relay is normally energized from the 48 volt DC from the RPS. When the voltage is removed by an automatic reactor trip signal, the relay is de-energized and releases the UV trip lever, which actuates the trip shaft, causing the breaker to unlatch from the closed position. The second method of tripping the trip shaft is by the shunt trip lever when the normally de-energized shunt trip (SHTR) coil is energized. When energized, the SHTR coil is powered from the 125 volt DC system used to close the reactor trip and bypass breaker closing circuits. For the reactor trip bypass breaker, the SHTR relay is energized only by a manual pushbutton. After the reactor trip bypass breaker is opened, then a contact in series with the SHTR relay opens to de-energize the coil. Thus, the SHTR relay is only momentarily energized. For the reoctor tri p bypass breaker, the SH T R relay is energized only by a manual pushbutton. After the reoctor trip bypass breaker is opened, then a contact in series with the SHTR relay opens to d&energizethecoil. Thus, the SHTR relay is only momentarily energized. Train A of the reactor protection system powers the UV and Shunt Trip coils for RTA and BYB, and train B powers the UV and Shunt Trip coils for RTB and BYA per Reactor Protection Functional System Diagram (FSD) A 181007, Figure F-1. Page: 46 of 200 1211412009 Previous NRC exam history if any: 012K2.01 012 Reedor Protecti on System K2 KnONIedgeof bus power suppliestothefoiloNing: (CFR: 41.7) K2.01 RPS channels, components, and interconnej:ions ..................... 3.33.7 Match justification: The 125V Aux Building DC busses supply the Reactor Trip Breakers and Bypass Breakers (RPS components). They provide power to the Reactor trip breakers for both closing power and one of the sources of power for tripping the breakers. To correctly answer this question, the power supplies to the Reactor Trip breakers must be understood, including the A train 125V Aux Building DC bus. Objective:

2. RELATE AND DESCRIBE the operation of the Reedor Trip Breakers and Reactor Trip Bypass Breakers to include the operation of the following

(OPS-40302F02):

Shunt Trip Coils Undervoltage Coils 1. RELATE AND I DENTI FY the operational characteristics including design features, capacities and protective interlocks for the following components associated with the Reedor Protecti on System (RPS) (OPS-522011 02):

Solid state protection system (SSPS) cabinets (A train/B train)
I nput rei ay cabi nets
Logi c cabi nets
Output rei ay cabi nets
Safeguards test cabi nets
Reedor tri p breakers
Reactor tri p bypass breakers Page: 47 of 200 1211412009 FNP Units 1 & 2 REACTOR PROTECTION SYSTEM A181007 3.2.7 Interface Requirements The STC shall interface with the SSPS and shall be supplied by qualified Class IE power from the 120 Vac vital power cabinets. (References 6.7.014,6.4.059,6.4.060,6.4.084) 3.3 REACTOR TRIP SWITCHGEAR TPNS Nos. Service QC11E004A-AB QC11E004B-AB (RTA, BYB) (RTB, BYA) 3.3.1 Basic Functions The reactor trip switchgear functions to switch power to or remove power from the control rod positioning equipment.

The switchgear opens the reactor trip and bypass breakers A and B on reactor trip causing the control rods to fall by gravity into the reactor core. 3.3.2 Functional Requirements The switchgear assembly shall consist of two low voltage metal enclosed switchgear sections. One section will contain two series connected reactor trip circuit breakers. The second will contain two bypass circuit breakers connected so that a bypass breaker parallels each reactor trip breaker. The bypass circuit breaker is used to bypass the reactor trip breaker for on-line testing of the latter with the reactor in operation. The system also includes two 260 volt line to line identical three phase Motor-Generator sets rated at 400 KV A, reverse current relay, generator output circuit breaker, a synchronizer, and a common ground relay. Each circuit breaker shall have provisions for locking it in the "Test" and "Disconnected" draw-out positions. The circuit breaker also includes positions for "Connected" and "Remove." (Reference 6.4.077) Interposing relays shall be used to isolate Train A from Train B wiring where it is necessary to parallel these circuits into a single output. Each circuit breaker shall be equipped with a 48 volt DC i Il.§1antaneous_ e a 125 V dc shunttrip device. (Reference ,.-----3-9 Rev. 0 FNP Units 1 & 2 REACTOR PROTECTION SYSTEM A181007 6.4.086) The Shunt Trip Attachment coil shall operate on 125 V dc and function as a backup for the undervoltage trip device. The first method of tripping the breaker (i.e., reactor trip or bypass breakers) is by a loss or drop of rated voltage to the Undervoltage Relay (UV). Ihe relay is DC from When the voltage is removed by an automatic reactor trip signal, the relay is de-energized and releases the UV trip lever, which actuates the trip shaft, causing the breaker to unlatch from the closed position. The second method of tripping the trip shaft is by the shunt trip lever when the normally de-energized shunt trip (SHTR) coil is energized. When energized, the SHTR coil is powered from the 125 volt DC systel!!. used to) , close the reactor trip and bypass breaker closing circuits. j1 I let fJ-/ For the reactor trip bypass breaker, the SHTR relay is energized only by a manual pushbutton. After the reactor trip bypass breaker is opened, then a contact in series with the SHTR relay opens to de-energize the coil. Thus, the SHTR relay is only momentarily energized. For the reactor trip breaker, the SHTR relay is energized by the closing of a contact associated with a shunt trip attachment relay (STA for 52/RTA and STB for 52/RTB). STA (STB) is energized from the RPS voltage to the UV trip coil of the 52/RTA (52/RTB). When the voltage is removed by an automatic reactor trip signal, the relay will de-energize, closing its contact to energize the shunt trip coil of 52/RTA (52/RTB). After the reactor trip breaker is opened, then a contact in series with the SHTR relay opens to de-energize the coil. Thus, the SHTR relay is only momentarily energized. 3.3.3 Design Transients The ambient design conditions are: 95% relative humidity and 40 deg. F to 120 deg. F temperature. (Reference 6.4.090) Also see Protection Features 3.3.7. 3-10 Rev. 0 BOARD MASTER AND ACTUATE TRAIN 'B' CONTROL { OUTPUT SLAVE RELAYS SAFEGUARDS <TRAIN 'B") ISOLA TION PROTECTION ANALOG PROTECTION SYSTEM SYSTEM --LOGIC SOLID STATE LOGIC ('J. 7300 SYSTEM) TRAIN B NUCLEAR INSTRUMENTATION SYSTEM / 1\ " INPUT DR FIELD CONTACTS { PROCESS CHANNEL_ SENSOR r IVl-----/ II )-) -------CHANNEL ---1 [ill-{ CHANNEL 'i @)--{ INPUT ;1 RELAYS <TYP.) COMPUTER j,-"DR" CABLE TO ROD DRIVE MECHANISM ROD CONTROL SYSTEM CONTROL I BOARD , __ /" -<", ( IBYPASS 1\ r { I t RKR. 'A' I INPUT CONTROL PROTECTION LOGIC n J r---f1 SYSTEM . . rr I' ¥\ TRAIN A i'1 \ L----------f __ ROD ISOLATION CONTROL ACTUATE TRAIN 'A' M-G SAFEGUARDS SET 1 ROD CONTROL M-G SET ! REACTOR TRIP REACTOR PROTECTION SYSTEM BOUNDARIES FIGURE F-l FNPUNIT 1 LOAD LIST A-S062S0 b V DF03 VI I ED04 LA13 IB 125V DC DIST PNL AB-139' D177082 (CONT'D) BKR TPNS DESCRIPTION SEE --PAGE lB-ll N1R15GOOO1C-N lF 4160V BUS BREAKER TEST CABINET lB-12 lA HOT SHUTDOWN PANEL .AUX.*RELAY CABINET F-56 ... >>> lB-13 Q1H21EOOO4-A lF 4160V BUS LOCAL CONTROL PANEL DIFFERENTIAL LOCKOUT RELAY CONTROL CIRCUIT lB-14 Q1R43EOOO1A-A lF BUS LOADING SEQUENCER CONTROL POWER TO: LOSP SEQ, LOAD SHEDDING, BKR CLOSE FAILURE & SEQ LOCAL ANNUN lB-15 Q1H21E0504-A lH 4160V BUS LOCAL CONTROL PANEL DIFFERENTIAL CONTROL CIRCUIT lB-16 Q1CllEOOO4B-AB TRIP SWITCHGEAR CONTROL POWER TO PASS BREAKER & REACTOR TRIP BREAKER lB.-J.7 Q1H21NBAFP2605A lA LOCAL HOT SHUTDOWN PANEL >>> F-57 ... A lG LOCAL HO.T SHUTDOWN PANEL >>> F-58 -A .. lB-18 Q1R43E0501A-A lH BUS LOADING SEQUENCER CONTROL POWER TO LOAD SHEDDING CONTROL CIRCUIT lB':"'19 Q1H25LOOO4-A .. 4A TERMINATION .. CABINET PANEL 4. REAR >>> F-59 lB-20 N1R15AOOO3-N lC 4160V BUS UNDERVOLTAGE AND UNDERFREQUENCY PROTECTIVE RELAYING lB-21* Q1H25LOOO6..;.A 6A TERMINATION CABINET PANEL 1 FRONT >>> F-60 lB-22 ------------ SPARE Isectf.doc Page F -54 Rev. IS

18. 012K6.03 001/FNP BANK/RO/MEM 3.1/3.5/N/N/3/CVRIY Unit 2 is at 100% power, and the following conditions occurred:

PT-455, PRZR PRESS, has failed off-scale HIGH.
NO Operator action has been taken. Which one of the following identifies the MINIMUM additional channels required to meet the RPS and ESF actuation logic to initiate any reactor trip or any safety injection on Pressurizer Pressure?

Reactor Trip Safety Injection A. 1 1 1 2 C. 2 1 D. 2 2 A -Incorrect. The first part is correct. The second part is incorrect, but plausible since if the failed instrument tripped all bistables in the fail safe condition it would be correct. It would also be correct after the applicable TS and procedure directed actions were complete (tripping all bistables), but the question specifies "assume no operator actions". Pressure SI is only on low pressure, and the instrument failing high does not automatically trip the low pressure bistable. The Reactor trip is on low or high pressure at this power level, and the high pressure condition would need only one more bistable in to cause a reactor trip. B -Correct. One more bistable on high pressure would cause a reactor trip, but the SI is actuated on low pressure only, so two more in the low pressure condition are required for an SI. C -Incorrect. First part is incorrect, but correct for a low pressure reactor trip. However, the High pressure reactor trip has one channel already tripped, and one additional channel will give a reactor trip signal. The second part is incorrect (see A). Both parts together are also plausible since confusion could cause choosing the exact opposite of the correct answer. o -Incorrect. The first part is incorrect, but plausible since for a low pressure reactor trip it is correct (see C). The second part is correct since the only SI PRZR pressure actuation is low pressure (see B). Page: 48 of 200 1211412009 Previous NRC exam history if any: 012K6.03 012 Reactor Protection System KG Knowledge d the effect of a loss or malfunction d the following will have on the RPS: (CFR: 41.7/45/7) K6.03Triplogiccircuits ................................................ 3.13.5 Match justification: A channel failure in one direction (failing high) causes a loss of the potential for meeting coincidence in the opposite direction (failing low) from that channel. This is one way of losing a trip logic circuit for one of the channels. This question presents a scenario where one of the 3 required trip logic coincidence circuits for Pressurizer pressure is lost, and knowledge of the effect on the RPS system is required to answer the question. Objective:

1. RECALL AND OESCRI BE the operation and function of the following reactor trip signals, permissives, control interlocks, and engineered safeguards actuation signals associated with the Reactor Protection System (RPS) and Engineered Safeguards Features (ESF) to include setpoint, coincidence, rate functions (if any), reset features, and the potenti aI consequences for improper condi ti ons to i ncl ude those items in the following tables (OPS-52201107):

Table 1, Reactor Trip Signals
Table2, Engineered Safeguards Features Actuation Signals
Table 5, Permissives
Table 6, Control interlocks Page: 49 d 200 1211412009 10/81200909:51 FNP-I-EEP-O B. Symptoms REACTOR TRIP OR SAFETY INJECTION Revision 37 I. The following are symptoms that require a reactor trip. if one has not occurred:

Reactor Trip 1. 2. 3. 4. 5. 6. 7. Source Range High Flux (If not blocked) Intermediate Range High Flux (If not blocked) Power Range High Flux. Low Setpoint (If not blocked) Power Range High Flux. High Setpoint Power Range High Positive Flux Rate OTlIT OPlIT Pressure NI -31. 32 (TSLB3 1-1.1-2) NI-35.36 (TSLB32-1,2-2) NI-41,42.43.44 (TSLB3 6-1.6-2.6-3.6-4) NI-41.42.43.44 (TSLB2 11-1.11-2.11-3. 11-4) NI Cabinets (TSLB2 12-1.12-2.12-3. 12-4) TI-412C.422C.432C (TSLB2 7-1.7-2.7-3) TI-412B.422B.432B (TSLB2 8-1.8-2.8-3) -2 19-1.19-2 19-3 Page 2 of 37 Setpoint Reference Surveillance Test Data Book for current S.P. 25% Rx Pwr 109% Rx Pwr +5%/2 sec. +credits 117% -penalties 110%-penalties 1865 psig (rate compensated) Coincidence 112 1/2 2/4 2/4 2/4 2/3 2/3 Gii) (Rx Pwr > 10%) 10/8/2009 10: 18 FNP-1-EEP-0 REACTOR TRIP OR SAFETY INJECTION Revision 37 II. The following are symptoms of a reactor trip: a. Any reactor trip annunciator lit. b. Rapid decrease in neutron level indicated by nuclear instrumentation.

c. All shutdown and control rods are fully inserted.

Rod bottom lights are lit. III. The following are symptoms that require safety injection. if one has not occurred: / ( SI Signv Instrumentation Setpoint Coincidence r-..... (TSLB) C lO r 2/3 1. ( PI-455 456.457 1850 psig pressure 1 2 17-1.17-2.17-3) (If not 15 c ed) ,----.. -2. Steam Line PI-474.484.494. 100 psid 1 steam line Differential PI-475.485.495. 100 psig less pressure PI-476.486.496 than other two (TSLB4 10-2.10-3.10-4. on 2/3 protection 11-2.11-3.11-4. sets 12-2.12-3.12-4. 13-2.13-3.13-4. 14-2.14-3.14-4. 15-2.15-3.15-4)

3. Low Steam Line PI-474.485.496 585 psig 2/3 pressure (TSLB4 19-2.19-3.19-4) (rate (If not blocked) compensated)

4. Containment PI-951.952.953 4 psig 2/3 pressure high (TSLBI 1-2.1-3.1-4)

5. Manual N/A N/A 112 IV. The following are symptoms of a safety injection:

a. Any SI annunciator lit. b. BYP & PERMISSIVE SAFETY INJECTION ACTUATED status light lit c. MLB-1 1-1 or MLB-l 11-1 lit d. HHSI flow greater than 0 gpm. c r Page 4 of 37 QUESTIONS REPORT for RO 2010 NRC EXAM SUBMITTAL 12-15-09 19. 013K2.01 002JNEW/RO/MEM 3.6/3.8JNJN/3/CVRlVER 5 EDITORIAL A loss of B Train Auxiliary Building 125V DC Bus has occurred on Unit 1. Which one of the following is the correct impact on B Train ESF Equipment control? The B Train SI actuated MOVs (1) automatically stroke upon an SI actuation, and B Train ESF pumps (2) be started in LOCAL at the HSP. (1 ) (2) A. will can will can NOT C. will NOT can D. will NOT can NOT Tuesday, December 15, 2009 6:38:08 AM 50 A -Incorrect.

The first part is correct (see B), however The second part is not correct, these breakers receive control power from B train DC and although there is an alternate control power that is placed into the circuit when in "LOCAL" at the HSP, it is also from B train DC. Plausible, since B train has alternate control power and SOP-36.6, CIRCUIT BREAKER RACKING PROCEDURE, has numerous cautions about an additional Control Power source for the B train ESF Pumps. The existance of Alternate Control power may cause confusion as to the ultimate source of the alternate control power. B -Correct. The B Train SI actuated (that would normally stroke upon an SI actuation signal) MOVs can be operated with or without DC power (a separate DC power source is provided to some of these valves if equipped with a disconnect for position indication only--and these valves do not stroke automatically following an SI actuation because of the "normal" position of that disconnect--ie MOV8808B). The control power for operation comes from the 600V AC supply for each MOV via transformer. The control power for operation of the B Train ESF breakers is supplied from B Train 125V Aux Bldg DC. Although equipped with an alternate control power source, that power is also supplied form B train DC on another breaker with a different cable run (for Appendix R concerns). C -Incorrect. The first part is incorrect (see B). Plausible, since some of these MOVs are equipped with a DC power supply for indication (MOV8808B). Further plausibility is provided from many solenoid operated valves auto stroke after SI, and they usually require DC power (although for the opening) The second part is also incorrect (see A). D -Incorrect. The first part is incorrect (see C). The second part is correct (see B). Page: 51 of 200 1211412009 Previous NRC exam history if any: 013K2.01 013 Engineered Safety Features Actuation System K2 Knowledge of bus power supplies to the following: (CFR: 41.7) K2.01 ESFAS/safeguards equipment control .............................. . 3.6* 3.8 Match justification: ESF equipment (pump) control requires DC for Pump breaker operation and breaker indication, even though the components themselves are powered from AC. ESF MOVs are powered from the 600V MCC AC and get control power for valve position indication from the same MCC AC source. To answer this question correctly knowledge of the power supplies for these ESF control functions is required. Wrote this question to intentionally stay away from 120V vital AC Instrumentation power due to potential overlap with other questions on this exam. Objective: 1 NAME AND IDENTIFY the Bus power supplies, for those electrical components associated with the Emergency Core Cooling System, to include those items in Table 4-Power Supplies (OPS-40302C04).

2. RELATE AND DESCRIBE the effect(s) on the Emergency Core Cooling System for a loss of an AC or DC bus, or a malfunction of the Instrument Air System (OPS-40302C06).

Page: 52 c:l200 1211412009 07/01109 15:27:43 CAUTION: Failure to deenergize all sources of DC control power while engaged in breaker racking could result in equipment damage AND severe personal injury OR death. Refer to Table 1 to determine IF breaker has alternate DC control power, AND IF additional action is required. A breaker in an ESF 4160V bus, 1I2F, 1I2G, 1I2K, 1I2L, 1I2J and 1I2H, cannot be left in the TEST position if the switchgear is required to be operable in Modes 1-4, unless the seismic modification has been implemented on both the breaker and the cubicle. In Modes 5, 6, and defueIed, the breaker may be left in the TEST position and the switchgear will still be operable. (See Precautions and Limitations.) PRIOR to racking 4160V Circuit Breaker, perform a visual inspection of visible control wires on the cubicle door to ensure:

no significant insulation damage is present.
visible wires are landed at terminations.

Removing Breaker to TEST Position 4.8.1 Insert racking lever on left side of fulcrum plate and attach to mating hole on breaker. (Refer to Fig. 2.) 4.8.2 Depress breaker release lever, this releases the breaker from the interlock bar. A. At the same time push down on the racking lever to disengage the main line contacts. B. When the plunger is out of the guide rail notch, the breaker release lever can be released. NOTE: It is possible to use the racking lever to pull the breaker to the TEST position, but caution should be used to ensure that the operator's hands do not slip off the lever, or that the handle does not slip off the lever. 4.8.3 Remove the racking lever from the breaker compartment. Version 55.0 07/01109 15:27:43 TABLE 1 FNP-0-SOP-36.6 TABLE 1 BREAKERS WITH ALTERNATE DC CONTROL POWER CAU'l7ION: . Failure to deenergizeaU sources of DC control power. while engaged in. breaker .. ackingcould result in equipment damage AND severe personal injury OR death. 1. Breakers in this table have two separate DC control power supplies. One DC supply feeds the breaker through the DC supply in the breaker cubicle. An alternate DC supply has been installed to provide a separate supply for components capable of being operated from the Hot Shutdown Panel (HSP). 2. Alternate DC power supplies are only energized when the handswitch on the HSP is place in LOCAL. 3. Remote DC control power fuses will be removed (replaced) from (in) breakers in the table in conjunction with securing (restoring) local DC control power OR the appropriate HSP control switch will be verified in the REMOTE position. COMPONENT BREAKER REMOTE CONTROL FUSE LOCATION Charging/HHSI Pump 1B/2B (B Train) DG07 HSP-C CharginglHHSI Pump I CI2C DG06 HSP-C CCW Pump 1A/2A DG04 HSP-C CCW Pump 1B/2B (B Train) DG05 HSP-C MDAFW Pump IB/2B DGIO HSP-C Pressurizer Heater Group 1B/2B EC11 HSP-C Page I of 1 Version 55.0 FNPUNIT 1 IG 4160VBUS BKR TPNS DG04 QIPI7MOOOIA-B DG05 Q 1 P 17MOOO lB-AB DG06 QIE21MOOOIC-B DG07 Q I E21 MOOO lB-AB DG08 QIR43A0502-B DG09 QIEIIMOOOlB-B DGlO QIN23MOOOIB-B DGII QIE13MOOOIB-B DG12 QIRl1BOOO6-AB DG13 DGI4 QIRl5BKRDGI4 DGl5 NIRIIA0502-N lsectg.doc LOAD LIST AB -121' DESCRIPTION A-S062S0 D177006 SEE PAGE IA STARTUP TRANSFORMER (ALTERNATE) <<< 4160V BUS>>> .1E .4160/600VSST (NQRMAL) >>>EE02 >>> IACCWPUMP IB CCW PUMP DISC SWITCH QIR18A00004B-B PUMP (B TRAIN SUPPL Y) IC CHARGING/HHSI PUMP L-I G-2 >>> lB CCW lB CHARGINGIHHSI PUMP DISC SWITCH QIRI8AOOOlB-B >>> lB CHARGING/HHSI PUMP (B TRAIN SUPPLY) IB DIESEL GENERATOR (EMERG) <<< 1 B RHRlLHSI PUMP IBAFWPUMP lB CTMT SPRAY PUMP IF 4160/600V SST DISC SWITCH >>> IF F -113 I 4160/600V SST >>> IF LOAD CENTER (B TRAIN SUPPLY) 4160V BUS>>:> J-l PT COMPARTMENT IB STARTUP TRANSFORMER (NORMAL) <<< ( Page G-l Rev. 13 FNP UNIT 1 DG03 EE05 DC DIST --_..---" TPNS lE-02 QIR16B0005-B lE-03 QIR15A0007-B lE-04 QIR16B0005-B iE-OS lE-06 QIR16 07-B lE-07 QIR15A0504-B NIR15A0509-N lE-08 QIR16B0007-B lsectg.doc LOAD LIST j) AB-121' DESCRIPTION A-506250 D177083 SEE PAGE lC 600V LOAD CENTER DC CONTROL POWER FOR INC BREAKERS EC02, EC07, EC08 & ECIO IG 4160V BUS DC POWER

DG03, DG07, DGll, DGI & DG13 IG 4160V BUS U/F TRIP AUX RELAYS (TRIP DG BKR DG08) lC 600V LOAD CENTER DC CONTROL POWER FOR FOR BREAKERS EC03, EC04, EC05, EC06, EC09, ECll, EC12, EC13 & EC14 G-51 I ...

....... PENETRATION PANEL >>> IE 600V LOAD CENTER DC CONTROL POWER FOR INC BREAKERS EE02, EE07 & EE12 IJ 4160V BUS DC CONTROL POWER FOR BREAKERS DJOl, DJ02, DJ03, DJ04, DJ06 & DJ07 IJ 4160V BUS U/F TRIP AUX RELAYS (TRIP DG BKR DJ06) IJ 4160V BUS BREAKER TEST CABINET IE 600V LOAD CENTER DC CONTROL POWER FOR FEEDER BREAKERS EE03, EE05, EE06, EE08, EE09, EEIO, EEll, EE13, EE14, EE15 Page G -49 Rev 6 FNPUNIT 1 DG03 EE05 LOAD LIST c __ _ 0 F II ------------ TPNS DESCRIPTION 1F-03 SPARE 1F-04 SPARE ?06250 D177083 SEE PAGE 1F-05 G22NAHR261 B-1B CATALYTIC H2 RECOMBINER DC CONTROL PANEL -N ANNUNCIATORS & ECV-1119, TCV-1114, ECV-1112 SOLENOIDS G-71 I IF-06

ENCAPSULATION SYSTEM CONTROL >>> 04 JUNCTION BOX FOR MSVR FLOODING SENSOR RELAYS 49-1, 49-2, 49-3 & LSX (2-3A FUSES) SPARE 1F-08 1F-11 lsectg.doc "Bn .. TRAIN. SAMPLE PANEL >>> VALVE CONTROL SHUTDOWN PANEL AUK RELAY CABINET .L.IV\,....n..&.I HOTSBUTDOWN PANEL SELECTOR ) Page G -69 G-73 I G-74 I G-75 I Rev. 6 FNPUNIT 1 DG03 EE05 LB14 LOAD LIST )CCOCAL HOT SHUTDOWN PNL SEL SWITCH BOX AB-121' D181664 FUSE TPNS DESCRIPTION RELAYS TR1-TR6 ON TRC-SVs 2213B-B & 2214B-B; PRZR PWR REL SVs 0444BA-B & BB-B AND ISO MOV 8000B-B; RWST MOV 0115D-B; CCW HX MOV 3047-B; "HSP SEL SW IN LOCAL" ALARM F8 52-DG06 LOCAL MODE DC CONTROL PWR FOR FEEDER CHG/HI HEAD SAFETY INJECTION PMP 1C F9 52-DG07 LOCAL MODE DC CONTROL PWR FOR FEEDER CHG/HI HEAD SAFETY INJECTION PMP 1B FlO 52-DG04 LOCAL MODE DC CONTROL PWR FOR FEEDER BKR TO PMP 1A F11 52-DG05 LOCAL MODE DC CONTROL PWR FOR FEEDER BKR TO CCW PMP 1B F12 52-DG10 LOCAL MODE DC CONTROL PWR FOR FEEDER BKR TO AFW PMP 1B F13 52-EC11 LOCAL MODE DC CONTROL PWR FOR FEEDER BKR TO PRZR HTR BACKUP GROUP 1B 1 ) lsectg.doc Page G -75 Rev. 6 j 2: :il :::1 :J ":" 41.(,,) 41(,) c 4'3 t MOC I I/. S D E L Time: r \04:26 PM ) 3 5 6 7 I" OPeN I' CLOSe _I' INDlCAiIOt-.!

'I ,." I C 11 33 .!:>3 "4 is ()V VlJV (1 ('2, _ Mea iT: --.k MCB r H w r F ijg it'f ¢I L-______________ " Mec. ---;;If-sp

40) -'3 -U-SEAL-jl-%Al

\\.1 IN -#-41.('>+ 41(0) -II-'51" -II-SP -,ff-sP +"'P o-J!-oSP o-Jl-osP <>-jf--<>sp 4--oSI> ¥ --U RI ?MCC. l\ Q_T\ AT' t!e A ru Ai, J!s B <1}\j I -J., Mes Cl-IAS'51S Y T GROUND Me£... d I , ( Title: C:\NRC EXAM SECURITY REQUIREO\HLT-33\1 HLT-33 Written\RO Oevelopment\Archive\0177614.cal Date: 10/r ""\009 6 -----, MODULI; --/.???7{l-IAr;<;IS 61l0UNO 3 ; 1 7 8 9 10 -i (NOTE 6) l LIMIT SWITCH CONTACT DEVELOPMENT CO>JT V;"'LV£ POSlTIO .... FUNCTION FULLY OP£N -FULLY I -S'fPA':>'" 'I',T Z -(NION. LI>T, c,\l., I ' D-I77414 SH, 1 3 GIl.. " OPlON 5 5 rAKE' .. = SPAIl.E 7 B CLOSED LII'II\\ 9 1== MC.I!> N..I>.RM 10 "'PAIlE II <;PAIlE IZ SPAne 13 1=== e,y PASS CKT, 14 r;PAIlI', .. Qil 16 '3PA12;;' _____ CLOSED NOT USED A-I77100 ,",H,I:l4 17 TORQUE SWITC.H HJTEA.Ruprs C.O>JTROL C.IRC.UIT IF" >;IEC.HA""'AL OvERLOA.D OC.C.uRS OUR\t.l6 Ct..OSI1,J6 C.YCLE. 18 OPEtoJlJ.J6 TORQUE 5W,TC.1-I I),JT£R.RUPTS COtJTROL CIRC.Uli U:: MECHAflJlCAL OVE.RLOA.D DURn"'60P£"'-II"-lG CYCL.E \71-1" eL'" ... mo OPE>I AUTO CO>>T t fl (LOSE! OPeN I A X B x 1. A x QI;MOIIO I-IAI-ICSWITC\\ CAl '5f'QING Q'aiUIlN ,0 AU,\,O 6)-QI\-H r;:-;:-:-11 '13.0 Time: ,. -"\,05:04 PM ! 12 13 DESCRIPi IOI'l REMARKS AUl(lUI\RY RE.LAY NOTES I. IIALII. IS '5\40Wl-l IN t=ULl'{ (L05E.D POSITION 'I. FOR M010R 1$ '5TAI'I.TE.1'I. '517..1:, 'SE.E DWG B-Il1'5;(D '514 T 19 :" CONTACT CLOSES ON SAfETY INJECTION ACTUATION SIGNAL, FOR RELAY AND CONTACT N°'S SEE EQIPMENT TABLE 4, REFER TO FUSE MANUAL A-IB1987 FDR FUSE SIZE AND TYPE. 5. COMPONENTS MARKED '.11TH AN (,,) ARE I,;lTHIN E.Q, SCOPE. FOR INSTALLATION DETAILS SEE E.Q, REFERENCE DRA'.IING LISTED BELo'.l, 6, SEE PDMS FOR LIMIT S'.IlTCH SETTING GUIDANCE. REFERENCE DRAWINGS ",af.l'aRAI. DE.TAll':> ANI) Non'S B-181612/36 CONN, DIAG HHSI TO RCS CL. ISO, MDV QIE21MDV8B03A-A I A B c D E L Title: C:\NRC EXAM SECURITY REQUIRED\HLT-33\1 HLT-33 Written\RO Development\Archive\D177614.cal Date: 10/f '?09 E -F '5!::R'IICE ",.,WE G HHSI TO RCS CL ISO MOV H -J 40)*3 L-______________________ -:>jf-SP -If--SEI>.L-If-'5EA\. l\.l IN + 41.(,)+ 41.(0) -II-'51' -If-SP -,ff-sf' .."jf-sP o-lf-osP o-I!->SP QIHIING I NI1;.21 I Nle21 I '5'5P1. SOCoIvI*e, zSI?>E>()A-A .. H"Sse03A'

_

.. -STAIn liP NUM6EI'I. 13'Z Time: r PM rKl SSP TERM NOS. & RELAY RELAY N" I CONTACT NO. S'OP'A SSF H{604 I TB603-1I-11 TBGO, VtI.>Cl. __ ' ____ I vtlt JrC. __ >OTE ____ I"' .... " __ .. TE ____ lvER.JrC. __ MTt IvtR . .c __ .. TE ____ ' ..... 111 __ D.\TE IVCR.NO. __ lOT[ ____ IVER.>Cl. __ DATE __ _ 1 T 2 T 3 I 4 I 5 I 6 I 7 Title: C:\NRC EXAM SECURITY REQUIRED\HLT-33\1 HLT-33 Written\RO Development\Archive\D177614.cal Date: 1 Olf -'\)09 Time PM . , SSP TERM NOS. & RELAY RELAY N" CONTACT NO. 1-1<604 S,<::-.P-A oJ-jP£---o SSP-A TB603-11 TBE.03-ll TC 1 2-41&1-10 .. QI:MoiE Io4ANDSWITCH 'l 61:.MCO CAl 404':;;"'Z.'I.'I.I-'(-AA3,A4 . SI'QING QS1UQN ,0 AUIO BLO'I', OIAGIZAM TERMINAL 'BLOCK NllMBER.'S 3 4-'i 6 7 8 4181-1, 41&1-lb 41BHI 41111-12 4161-14 REFERENCE DRAWINGS GEt-.lSRAL DI:TAll':> AND B-IB1612/36 CONN. DrAG HHSI TO RCS CL. ISO. MOV QIE21MOVBB03A-A D-IB1900/14 INSTALLATION DETAILS FOR LQ. LIMITORQUE MOV'S . @ DI776141 [yVY2005 1 JLO -031 Southern Company Services. Inc. for ALABAMA PO\JER COMPANY -I E F G H fOf use by emplOyees of. outhoriz:fcOntroctors of. the sobsidiories of The So..them Com",,,), ....... 1>0 .... ..... "'tn'ulion. copying. PI<I1.<C' J.M. FARLEY NUCLEAR PLANT -UNIT NO.1 J 6_ 0' 'od"" .. of ""I pori ... """,;, prohibited. ELEMENTARY DIAGRAM ,y BY ,y J'( CKCKD HMJ IaICKD --- MOTOR OPERATED VALVE MDV8803A-A I I I I I I I I I I I I I I I I I I I I I I JLOIRlPIJURI = .'" 8-14-72 "'u: NONE JIM _nr.T£ 'tI1:R.1fl --" ----V[R,)(l. __ OATE ----vtR.'" __ OA" ----vt."" 11-30-07 """'vtJ ""'" 1 .., _ '><En I vtR. 1.0 InKRSEDtS D-177614 7 8 9 10 11 12 13 Title: C:\NRC EXAM SECURITY REQUIRED\HLT-33\1 HLT-33 Written\RO Oevelopment\Archive\0177614.cal FNPUNIT 1 DF03 ED14 IV 600/208V MCC (CONT'D) BKR TPNS FUN4 ------------ FUN5 Q1E15MOV3362A-A FU02 Q1N23MOV3764D-A FU04 Q1N23MOV3764F-A FU05 Q1E21LCVl15B-A FUP2L Q1N12MOV3406-AB FUP4L ------------ FUR2 Q1E21MOV8886-A FUR3 Q1E21MOV8803B-AB FUR4 Q1EllMOV8811A-A FUR5 Q1E14MOV3660-A FUS2 Q1P16MOV3019A-A FUS3 Q1P16MOV3019B-A FUS4 Q1P17MOV3094B-A FUS5 ------------ FUT2 Q1N23MOV3209A-A FUT3 Q1N23MOV3210A-A 1 sectf.doc LOAD LIST A-506250 AB -139' B177556-19 DESCRIPTION SEE PAGE SPARE PENETRATION ROOM TO PENETRATION ROOM FILTER MOV AUX FEEDWATER TO STEAM GENERATOR MOV AUX FEEDWATER TO STEAM GENERATOR MOV RWST TO CHG PMP SUCTION ISOLATION I TURBINE DRIVEN AUX FEED PUMP MOV 1U 600/208V MCC XFMR >>> 1U MCC 208V F-I09 SECTION >>> DISC SWITCH Q1R18B029-A >>> HHSI TO RCS HOT LEG MOV HHSI TO RCS CL ISO MOV CTMT SUMP OUTLET MOV CTMT AIR SAMPLE MOV CTMT COOLER SERVICE WATER INLET MOV CTMT COOLER SERVICE WATER INLET MOV SPENT FUEL POOL HX INLET MOV SPARE AUX FEEDWATER PUMP SERVICE (MD) INTAKE MOV AUX FEEDWATER PUMP SERVICE (MD) INTAKE MOV , / ) Page F -107 Ver. 56.0

20. 013K5.01 001/NEW/RO/C/A

-2.8/3.21N/N/3/CVRN Unit 2 was operating at 100% power, and the following conditions have occurred:

PT-950, CTMT PRESS, has failed.
PT-950, HI-3 bistable, is in the BYPASS condition.
Subsequently, the 20 vital panel has become de-energized.

If a Large Break LOCA occurs, which one of the following describes:

1) the number of channels of Hi-3 bistables which will be actuated and 2) the number of trains of containment spray (CS) that actuate automatically? 1) Two channels ONLY will be actuated.

2) One train ONLY will actuate. B. 1) Two channels ONLY will be actuated.

2) Two trains will actuate. C. 1) Three channels will be actuated.

2) One train ONLY will actuate. o. 1) Three channels will be actuated.

2) Two trains will actuate. Page: 53 of 200 1211412009 A -Correct. The channel I bistable is bypassed and won't actuate. The channel IV Bistable won't actuate since it is deenergized by the loss of 1 D vital 120V AC, and it is an energize to actuate bistable.

Even though the coincidence for CS actuation would still be met with channel II & III, and both trains of SSPS would get the signal to actuate both trains of CS, the slave relays are deenergized in SSPS train B due to the loss of 1 D vital 120V AC panel. This would prevent Train B CS from actuating. B -Incorrect. The first part is correct (see A). The second part is incorrect, but plausible since both trains of SSPS get a signal to initiate CS, even with a loss of 1 D 120V vital AC panel. The master relays call for an actuation on Both trains, but on B train the slave relays don't have power to start the loads and operate the valves for the ESF actuations. C -Incorrect. The first part is incorrect, but plausible. For most bistables, when they lose power they deenergize to actuate. Containment spray is an exception to this general rule. Channel IV is thus deenergized and will NOT actuate. Examinee could also not realize that the bypass function (which prevents the bistable from actuating) is opposite the usual trip bistable function (which causes the bistable to trip) for a loop in maintenence. This would cause this choice to be selected. The second part is correct (see A). D -Incorrect. Both parts are incorrect (See C & B). FSD: A181007 REACTOR PROTECTION SYSTEM 2.2.2 The RPS system is housed in two physically and electrically independent equipment trains (Train II A" and Train II B"), typically referred to as the Solid State Protection System (SSPS) cabinets. (Reference 6.7.003) 2.2.3 Any singlefailurewithin the RPS system (sensor channel or actuation train) shall not prevent the redundant system actuation. On loss of channel or train poNer the bistable shall be tripped. T he only exception to the loss of channel or train poNer causing the bistable to trip is for Containment Spray and Containment Phase B I SJlation where the bistables must energize to actuate. (References 6.1.002, 6.1.41, 6.7.014,) 2.2.61 nstrument channels shall be poNered from four separate independent AC instrument distribution panels. These panels shall be fed from four separate and independent Class 1E inverters. (References 6.1.023,6.4.081,6.4.091,6.7.014,6.7.016) 2.2.15 Except as noted bel ow, all reactor tri p and safeguards actuati on channels shall be pi aced in the trip mode when the channel is out of service for any reason. The reactor trip and safeguards actuation circuits noted belON be administratively bypassed for mai ntenance on a si ngl e channel. 1. Source range hi gh neutron fl ux tri p 2. I ntermedi ate range hi gh neutron tri p 3. High 3 contai nment pressure actuati on of contai nment spray LOAD LIST: A-506250, PageG-39, 1D 120V Vital AC Dist PNL. Page: 54 of 200 1211412009 Previous NRC exam history if any: 013K5.01 013 Engineered Safety Features Actuation System K5 Knowledge of the operational implicationsofthefoilONing concepts as they apply totheESFAS: (CFR: 41.5/45.7) K5.01 Detinitionsof sctety tran end ESF chennel ........................... 2.83.2 Match justification: To answer this question correctly, knowledge is required of what constitutes a safety train, an ESF channel, and the operational implications of each must be understood. Objective:

1. RECALL AND DESCRI BE the operation and function of the followi ng reactor trip signals, permissives, control interlocks, and engineered safeguards actuation signals associated with the Reactor Protection System (RPS) and Engineered Safeguards Features (ESF) to i ncl ude setpoi nt, coi nci dence, rate functi ons (if any), reset features, and the potenti aI consequences for improper condi ti ons to i ncl ude those items in the followi ng tables (OP&52201107):

Table 1, Reactor Trip Signals
T abl e 2, Engi neered Safeguards Features A ctuati on Si gnal s
Table 5, Permissives
Table 6, Control interlocks Page: 55 of 200 1211412009 FNP Units I & 2 CONTAINMENT SPRAY . JJ SETPOINT High-? {/'i\ Manual NIA REACTOR PROTECTION SYSTEM TABLE T-4 -ENGINEERED SAFEGUARDS ACTUATION SIGNALS by CONTAINMENT SPRAY ACTUATION COINCIDENCE 2/4 High-3 containment pressure signals 2: setpoint INTERLOCKS

& BLOCKS None JP i?C'/'v k / PROTECTION PROVIDED FOR Protects containment for a loss of coolant or steam line break inside containment Prevents over pressurization of containment structure and subsequent building rupture -214 switches -Consist of four -Containment spray Operator discretion f actuation is manually momentary switches in two groups reset by depressing both train A and train B reset push buttons onMCB T4-3 MODES OF OPERATION 1,2,3 1,2,3,4 A-18lO07 FSD SECTION 2.4 2.7.1 2.7.2 Fig. 2 Sht. 8 2.7.1 Fig. 2 Sht. 8 Rev. 0 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABI L1TY Farley Units 1 and 2 ESFAS Instrumentation B 3.3.2 b. Containment Spray -Automatic Actuation Logic and Actuation Relays (continued) the use of the Manual Initiation Switches. Automatic Actuation Logic and Actuation Relays must be OPERABLE in MODE 4 to support system level Manual Initiation. In MODES 5 and 6, there is insufficient energy in the primary and secondary systems to result in containment overpressure. In MODES 5 and 6, there is also adequate time for the operators to evaluate unit conditions and respond, to mitigate the consequences of abnormal conditions by manually starting individual components.

c. Containment Spray -Containment Pressure -High 3 This signal provides protection against a LOCA or an SLB inside containment.

The transmitters (dip cells) and electronics are located outside of containment with the sensing line (high pressure side of the transmitter) located inside containment. Thus, the transmitters will not experience any adverse environmental conditions and the Trip Setpoint reflects only steady state instrument uncertainties. This Function requires the bistable output to energize to perform its required action. It is not desirable to have a loss of power actuate containment spray, since the consequences of an inadvertent actuation of containment spray could be serious. Note that this Function also has the inoperable channel placed inEYpass rather than trip to decrease the probability of an inadvertent actuation. The Containment Pressure High 3 instrument Function consists of a two-out-of-four logic configuration. Since containment pressure is not used for control, this arrangement exceeds the minimum redundancy requirements. Additional redundancy is warranted because this Function is energize to trip. Containment Pressure -High 3 must be OPERABLE in MODES 1, 2, and 3 when there is sufficient energy in the primary and secondary sides to pressurize the containment following a pipe break. In ( continued) B 3.3.2-14 Revision 0 FNP Units 1 & 2 REACTOR PROTECTION SYSTEM A-181007 injection is required and containment spray may be necessary to ensure the integrity of the containment building and limit offsite dosage. Actuation signals for these protective functions are provided by containment pressure transmitters. If 2/3 high-l containment pressure channels sense a pressure greater than 4 psig, then safety injection will be initiated. (References 6.1.002, 6.4.007,6.4.015,6.4.036)

2. Containment High-2 Pressure If containment pressure increases to 16.2 psig on 2/3 containment pressure channels, a main steam line isolation signal is generated.

This signal is installed to isolate a leaking steam line and prevent further pressure increases in containment. (References 6.1.002, 6.4.007,6.4.015,6.4.036)

3. Containment High-3 Pressure The high-3 containment building pressure detection circuitry functions to initiate a Phase B containment isolation and actuate the containment spray system. Spraying is necessary to prevent overpressurization of the containment structure and maintain structural integrity to limit the offsite dosage. The high-3 pressure setpoint is 27 psig. An actuation signal occurs when 2/4 containment pressure channels exceed their setpoints.

These are the only circuits which are energized to actuate, because inadvertent spray actuation is not desirable. The containment spray actuation shall utilize all four pressure channels in 2/4 coincidence logic. Only one channel should be bypassed at a time. An additional protection channel degraded by single failure criteria should not render the spray actuation inoperable. The 2/4 coincidence logic which is utilized for containment spray ensures the availability of the minimum protection channels for actuation. It meets the requirement of IEEE-279. Moreover, the reliability of the circuit is established by periodic testing. (References 6.1.002, 6.4.007, 6.4.015, 6.4.036) 4. Low Pressurizer Pressure A steam break accident or a LOCA will cause a decrease in pressurizer pressure due to the outsurge of water from the pressurizer. The outs urge will be caused by either the water loss from the LOCA or the rapid cooldown produced by the steam break. Safety injection will be initiated by low pressurizer pressure of 1850 psig, as sensed by 2/3 pressure detectors. This meets the 2-29 Rev. 10 I FNP Units 1 & 2 ACTUATION SIGNAL High-l containment pressure PB951B PB952B PB953B Manual SETPOINT 4 psig N/A REACTOR PROTECTION SYSTEM TABLE T-4 -ENGINEERED SAFEGUARDS ACTUATION SIGNALS SAFETY INJECTION ACTUATION (CONTINUED) COINCIDENCE 2/3 High-l pressure signals ::: setpoint 112 Momentary switches INTERLOCKS & BLOCKS None None T4-2 PROTECTION PROVIDED FOR Loss of coolant or steam line break within containment Operator discretion MODES OF OPERATION 1,2,3 1,2,3,4 A-lSlO07 FSD SECTION 2.4 2.7.1 2.7.2 Fig. 2 Sht. S 2.7.1 Fig. 2 Sht. S Rev. 0 10/8/2009 13: 10 FNP-2-EEP-0 REACTOR TRIP OR SAFETY INJECTION Revision 33 \ II. The following are symptoms of a reactor trip: a. Any reactor trip annunciator lit. b. Rapid decrease in neutron level indicated by nuclear instrumentation.

c. All shutdown and control rods are fully inserted.

Rod bottom lights are lit. III. The following are symptoms that require safety injection. if one has not occurred: SI Signa:-) Instrumentation Setpoint "-(TSLB) l. Pressurizer PI-455.456.457 1850 psig pressure low (TSLB2 17-1.17-2.17-3) (If not blocked) 2. Steam Line PI-474.484.494. 100 psid Differential PI-475.485.495. pressure PI-476.486.496 (TSLB4 10-2.10-3.10-4. 11-2.11-3.11-4. 12-2.12-3.12-4. 13-2.13-3.13-4. 14-2.14-3.14-4. 15-2.15-3.15-4)

3. Low Steam Line PI-474.485.496 585 psig pressure (TSLB4 19-2.19-3.19-4) (rate (If not blocked) compensated)

4. Containment PI-951.952.953 4 psig pressure high (TSLB1 1-2.1-3.1-4)

,-5. Manual NIA N/A IV. The following are symptoms of a safety injection:

a. Any SI annunciator lit. b. BYP & PERMISSIVE SAFETY INJECTION ACTUATED status light lit c. MLB-1 1-1 or MLB-1 11-1 lit d. HHSI flow greater than 0 gpm. Coincidence 2/3 1 steam line 100 psig less than other two on 2/3 protection sets 2/3 2/3 1/2 FNP Units 1 & 2 REACTOR PROTECTION SYSTEM A-181007 Trip/Actuation Accuracy -This definition includes comparator accuracy, channel accuracy for each input, and rack environmental effects. This is the tolerance expressed in process terms (or percent of span) within which the complete channel shall perform its intended trip/actuation function.

This includes all instrument errors but no process effects such as streaming. 2.2 GENERAL FUNCTIONAL REQUIREMENTS-REACTOR TRIP SYSTEM, ENGINEERED SAFETY FEATURES ACTUATION SYSTEM. 2.2.1 The Reactor Protection System acts to limit the consequences of ANSI Condition II events (e.g. loss offeedwater, etc.). Minimum departure from nucleate boiling ratio (DNBR) shall not be less than the designed limit DNBR as a result of any anticipated transient or malfunction for Condition II events and as such shall meet 95/95 criterion. That is, departure from nucleate boiling will not occur on at least 95 percent of the limiting fuel rods at 95 percent confidence. (DNBR design limit for vantage 5 fuel is 1.2411.23 for the typical and thimble cells and 1.2511.24 for LOPAR fuel for the typical and thimble cells respectively). Rod linear Power density shall not exceed the rated design value (22.4 KW/ft.) and the stress limits of the reactor coolant system as specified for Condition II events (2735 psig) shall not be violated. Release of radioactive material for any Condition III fault shall not be sufficient to interrupt or restrict public use of those areas beyond the exclusion radius and shall not exceed the guidelines of 10 CFR 20. For any Condition IV fault, release of radioactive material shall not result in an undue risk to public health and safety nor shall it exceed the guidelines of 10 CFR 100, "Reactor Site Criteria." The Engineered Safety Feature Actuation System, in addition to Reactor Trip, limits the consequences of ANSI Condition III events and mitigates ANSI Condition IV events. Table T-l summarizes the ANS classification of faults and safety analysis approach as outlined in FNP-FSAR-Chapter

15. Most of the analyzed events/transients can be detected by one or more protection functions.

The various Safety Analyses take credit for most of the protection functions. Those functions for which no credit is taken in the analyses are still required to be operable to enhance the overall reliability and diversity of the protection system. This design shall meet the requirements of IEEE-279-1971. (References 6.1.008, 6.1.031, 6.7.013, 6.7.039) 2-3 Rev. 6 FNP Units 1 & 2 2.2.2 2.2.3 2.2.4 2.2.5 2.2.6 2.2.7 2.2.8 2.2.9 2.2.10 2.2.11 REACTOR PROTECTION SYSTEM A-181007 The RPS system is housed in two physically and electrically independent equipment trains (Train "A" and Train "B"), typically referred to as the Solid State Protection System (SSPS) cabinets. (Reference 6.7.003) Any single failure within the RPS system (sensor channel or actuation train) shall not prevent the redundant system actuation. On loss of channel or train power the bistable shall be tripped. The only exception to the loss of channel or train power causing the bistable to trip is for Containment Spray and Containment Phase B Isolation where the bistables must energize to actuate. (References 6.1.002, 6.1.41, 6.7.014,) Actuation shall be automatic when the limits of the monitored parameters are exceeded. While one train is in test, the redundant train shall be capable of performing Reactor Trip/ESF AS actuation. (References 6.1.002,6.7.014,6.7.015) The Reactor Protection System shall have provisions in the control room for manually initiating the reactor trip or actuating engineered safety features systems. (References 6.1.002, 6.7.014) Instrument channels shall be powered from four separate independent AC instrument distribution panels. These panels shall be fed from four separate and independent Class IE inverters. (References 6.1.023, 6.4.081,6.4.091,6.7.014,6.7.016) Auxiliary devices that are required to operate on an ESF AS actuation to support train-related functions, shall be supplied from the same distribution panel to prevent the loss of electric power in one protection set from causing the loss of equipment in the redundant protection set. (References 6.1.023, 6.4.081, 6.4.091) Each distribution panel shall have access to its respective inverter and standby power supply. (References 6.1.023, 6.4.081, 6.4.091) A protective action at the system level, once initiated, shall go to completion. Actuation is sealed-in until manually removed from operation. (Reference 6.7.014) Each RPS actuation shall be alarmed and annunciated in the control room. (Reference 6.7.014) Protection interlocks and bypasses shall be designed to meet the requirements ofIEEE 279-1971 Sections 4.12 through 4.14. (Reference 6.7.014) 2-4 Rev. 4 FNP Units 1 & 2 2.2.12 2.2.13 2.2.14 2.2.15 2.2.16 2.2.17 REACTOR PROTECTION SYSTEM A-18!007 The RPS shall be capable of testing at power and shall follow the guide lines of Regulatory Guide 1.22, IEEE 338-1971, and IEEE 279-1971 Section 4.10. (References 6.7.014, 6.7.015, 6.7.032) ESF shall be tested either in "GO TEST" mode or in "BLOCK" test mode. In "GO TEST" mode the ESF device is operated, or equipment alignments for special operation are performed. In the "BLOCK" test mode, where the end device test would cause plant upset, the end device actuation is blocked while the "ACTUATION" signal is verified by continuity check. Typical examples are Feedwater control valves, steam line isolation valves, RCP breaker test, etc. (References 6.7.003, 6.7.011, 6.7.032) Electrical or mechanical interlocks and bypasses on safety-related equipment, when initiated manually or automatically, shall be continuously indicated in the main control room. No more than one train or channel may be bypassed at one time. (Reference 6.7.014) Except as noted below, all reactor trip and safeguards actuation channels shall be placed in the trip mode when the channel is out of service for any reason. The reactor trip and safeguards actuation circuits noted below may be administratively bypassed for maintenance on a single channel. 1. Source range high neutron flux trip 2. Intermediate range high neutron trip 3. High 3 containment pressure actuation of containment spray (References 6.1.004,6.1.011,6.1.022,6.4.007,6.7.014,6.7.045,6.7.046) Channel independence shall be required throughout the system, extending from the sensor through the devices actuating the protective function. Redundant logic system cabinets shall be maintained and separated from the analog channels. Reactor trip and ESF AS analog circuits may be routed in the same raceway if the circuits have the same power supply and sensor protection channel set. Cabinet separation criteria shall be verified by tests. (References 6.1.011, 6.1.042, 6.1.046, 6.4.081, 6.4.091, 6.7.001, 6.7.002,6.7.003,6.7.014,6.7.054) The system shall have functional diversity. As an example, for a coolant accident, a safety injection signal can be obtained manually or by automatic initiation from two diverse parameter measurements. These are: 1. Low pressurizer pressure 2-5 Rev. 4 FNPUNIT 1 DG03 EE05 LB06 LOAD LIST ID 120V VITAL AC DIST PNL AB-121' BKR TPNS DESCRIPTION INSTRUMENTATION 10-02 Q1H11NGNIS2503D-4 10 NUCLEAR INSTRUMENTATION CHANNEL 4 10-03 Q1H11NGPIC25050-4 PROCESS I&C PROTECTION CAB >>> 10.,...04 Q1H11NGPl:C2505H-4. PROCESS* I&C CONTROL.CAB#8 10-05 N1C56LOO01B-N 1B INCORE T 10 .. 06 N1.G22NBWPP2603C-N WASTE GAS PANELS SYSTEM SYSTEM A-506250 D177025 CABINET CABINET SEE PAGE CHANNEL 4 G-40 CHANNEL 4 >>> G-41 BOX >>> G-43 10-07 Q1H11NGSSP2506K-A RAIN A SOLID STATE PROT SYSTEM INPUT CAB CH 4 10-08 Q1H11NGSSP2506G-B TRAIN B SOLID STATE PROT SYSTEM INPUT CAB CH 4 10-09 ------------ SPARE 10-10 N1H11NGAR2506E':'B AUX .RELAY RACK TRAIN B >>> MAIN . BOARD .SECTION. A>>> 1A CONTROL BOARD DEMULTIPLEXER CABINET TRAmB r.n.l: * .I:a.L.I >>> >>> lsectg.doc Page G -38 G-44 G-45 G-46 G-47 G-48 Rev 6 I I FNPUNIT 1 DG03 EE05 LOAD LIST AB-l2l' BKR TPNS DESCRIPTION A-S062S0 D177025 10-16 Q1H11NGSSP25061 TRAIN B SOLID STATE PROTECTION SYS -B '---------- --. 10-17 ------------ SPARE 10-18 Q1H11NGSSP2506J TRAIN B SOLID STATE PROTECTION SYS SAFEGUARDS -B TEST CAB f ,+ "-VI I ) ) " ) 1 sectg.doc Page G -39 Rev 6

21. 015AK2.10 001/NEW/RO/C/A

-2.Bl2.8/N/N/2ICVRIY Unit 1 is at 20% power and conditions are as follows: At 1000: 1A 1B 1C

RCP amps: 670 680 690 At 1005: 1A 1.!2 1C
RCP amps: 670 680 0
EF3, 1 C RCS LOOP FLOW LO OR 1 C RCP BKR OPEN, is in alarm. Which one of the following describes the expected indications on 1A RCS LOOP and 1 C RCS LOOP flow rates at 101 O? 1 A RCS LOOP Flow rate 1 C RCS LOOP Flow rate A. 105% and stable 0% and stable 105% and stable 10% and stable C. 100% and stable 10% and stable D. 100% and stable 0% and stable Page: 56 d 200 1211412009 A -Incorrect.

The first part is correct (See B). The second part is incorrect, but plausible, since the RCP amps are 0, and it is tripped as indicated by the bkr light and amps. If it weren't for reverse flow caused by the discharge pressure of the other two pumps 0% would be correct. B -Correct. Each of the two loops with forced flow provide some backflow through the tripped pump (approx 5% each for a total of 10%). The tripped pump has an indicated flow (approximately 10%) due to the flow indicator sensing a positive value of flow >0, even though the direction of flow is reversed. C -Incorrect. The flow in the 1A & 1 B loops is greater than 100%, but this is plausible since the amps in the 1A & 1 B pumps are unchanged. The increased flow is due to decreased resistance to flow downstream of these pumps which is why flow increases with no increased amps to the pump motors. The second part is correct (see B). o -Incorrect. Both parts incorrect but plausible. This would seem to be indicated by the amps and the fact that the 1 C pump is not pumping any flow due to being tripped by annunciator indication. However the piping system of the RCS allows back flow into the C loop in this condition and it is indicated on all three loop flowmeters. Ran on simulator laptop to verify flows. No technical document was found which stated this characteristic in writing. Loss Of Reactor Coolant Flow, OPS-62520D OPS-52520D, Student Text-Version 2, listed this. Cvr 8-4-09 Previous NRC exam history if any: 015AK2.10 015 Reactor Coolant Pump Malfunctions AK2. Kncmledge of the interrelations between the Reador Coolant Pump M alfundions (Lex:;s of RC Flc:m) and thefollc:ming: (CFR 41.7/45.7) A K2.1 0 RCP i ndi cators end control s . . . . . . . . . . ............................. 2.8* 2.8 Match justification: This question provides indications which accompany a RCP trip and must be recognized as such. The knowledge of the interrelations between the RCP loss of flow and the flow indicators (and what to expect the RCP flow indicators to read after a RCP trip) must be used to obtain the correct answer. Objective:

2. RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the components associated with the Reactor Coolant System (RCS) to include the components found on Figure 1, Reactor Coolant System (OPS-40301A02).

Page: 57 of 200 1211412009 will effectively increase the unaffected SG's steaming rate and power output to compensate for the reduced steaming rate from the affected SG. During the transient, the affected SG main feed regulating or bypass feed regulating valve controller(s) is in automatic. Initially, the regulating valve will go shut in response to the reduced steam flow from the affected SG. This causes SG level in the affected SG to begin to fall. Contributing to the drop in SG level is the phenomenon of SG shrink. Shrink is caused by the density change in the tube section of the SG, which occurs as a result of the drop in temperature on the primary side of the affected SG. The drop in indicated level causes the feed valves to reopen fully in an attempt to bring SG level back up. Overfeeding of the affected SG can occur, which could lead to a turbine trip and SG feed pump (SGFP) trips followed by a reactor trip. The turbine trip and SGFP trip occur at 82% SG level (P-14). To minimize the effects a loss of coolant flow has on the affected SG level, the operator is instructed to take manual control of the affected SGs feed regulating valves. Another concern on any loss of coolant flow situation is pressurizer pressure control. Automatic control of pressurizer pressure will be affected due to the loss of spray flow if the loss of coolant flow occurs in loops A and/or B. If only one loop is involved, the affected loop's spray valve controller should be placed in MANUAL and the valve closed to prevent spray flow from the unaffected loop bypassing the pressurizer. If both A and B loops are affected, auxiliary spray flow should be utilized if normal letdown is available. The loop flow indications observed by the operators would be as follows: For the affected loop, flow would slowly decrease to 0 and then return to approximately 10%; for the unaffected loops, the flow should increase to approximately 1 05% (each loop). The flow indication in the idle loop occurs as flow stops and then begins again in the reverse direction. Since flow rates in the RCS loops are derived from the differential pressure felt in an elbow in each loop, any flow at all will be indicated, regardless of the direction. The indication observed in the two loops with the running pumps is due simply to the pumps in those loops picking up a small portion of the flow lost in the idle loop. 2 OPS-6252001525200 -Version 2

22. 022A 1.01 001/NEW/RO/MEM 3.6/3.7/N/N/3/CVRIY Unit 2 is at 100%, and the following conditions occurred:

At 1000:

The Containment Cooling system is in the normal mode of operation per SOP-12.1, Containment Air Cooling System.
Containment temperature is slowly rising. At 1100:
The crew has configured the containment cooling system per SOP-12.1.
The emergency service water from CTMT coolers; MOVs 3024A, B, C and 0 are OPEN lAW SOP-12.1.

Which one of the following identifies the: 1) MINIMUM temperature at which a Technical Specification action statement must be entered for Tech Spec 3.6.5, Containment Temperature, and 2) the speed of the Containment Cooling Fans lAW SOP-12.1? A. B. D. Page: 58 of 200 (1 ) (2) FAST SLOW FAST SLOW 1211412009 A -Incorrect. The 110°F is incorrect for Ctmt temp limit, but is plausible, since 110°F is the temperature per the CTMT HI TEMP alarm ARP-1.2, BB3, to start all ctmt dome recirc fans in fast speed. The second part is correct per SOP-12.1 step 4.1.6 note. B -Incorrect. The first part is incorrect (See A). The second part is incorrect but plausible, since Slow is the speed that the fans automatically shift to in a very high temperature LOCA environment. However, it is due to the humidity rather than the heat that they are shifted to slow to protect the fans. Under normal conditions, the lower humidity allows fast speed operation to remove more heat from containment. C -Correct. 120 degrees F is the TS 3.6.5 limit. Fans must be operated in fast per note prior to step 4.1.6 normally to maintain ctmt less than this limit per SOP-12.1, Step 4.1.9-4.1.11, ver. 37.0. o -Incorrect. The first part is correct (See C). The second part is incorrect (See B). ARP-1.2, 883, CTMT AI R TEM PHI, Version 44.0 OPERA TOR ACTION 5. IF contai nment average ai r temperature is greater than 110° F, TH EN veri fy contai nment dome recirc fans in service on fast speed. Previous NRC exam history if any: 022A1.01 022 Contai nment Cool i ng System A 1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) asariated with operating the CCS controls induding: (CFR: 41.5/45.5) A1.01 Containmenttemperaure ......................................... 3.63.7 Match justification: Question asks what the TS containment temperature limit is, and which controls of the Containment cooling system must be operated (fast or slow speed fans) to prevent exceeding the containment temperature limit. Objective:

7. DEFINE AND EVALUATE the operational implications of normal/abnormal plant or equipment conditions associated with the safe operation of the Containment Ventilation and Purge System components and equi pment, to i ncl ude the foil owi ng (OPS-40304A07)

Page: 59 ci 200

Normal control methods Abnormal and Emergency Control Methods Automatic actuation including setpoint (example SI, Phase A, Phase B, MSLlAS, LOSP, SG level) Protecti ve i sol ati ons such as hi gh flow, low pressure, low level i ncl udi ng setpoi nt Protecti ve i nterl ocks 1211412009 3.6 CONTAINMENT SYSTEMS 3.6.5 Containment Air Temperature Containment Air Temperature 3.6.5 LCO 3.6.5 Containment average air temperature shall be :<::; 120°F. APPLICABILITY:

MODES 1, 2, 3, and 4. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment average air A.1 Restore containment 8 hours temperature not within average air temperature limit. to within limit. B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. B.2 Be in MODE 5. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify containment average air temperature is within 24 hours limit. Farley Units 1 and 2 3.6.5-1 Amendment No. 146 (Unit 1) Amendment No. 137 (Unit 2) 09/24/06 13:00:16 FNP-2-ARP-l.2 LOCATION BB3 OPERATOR ACTION (continued)

4. IF possible, THEN start additional fan coolers to insure that the containment average air temperature does not exceed 120°F. 5. IF containment average air temperature is greater than 110°F, THEN verify Containment Dome Recirc Fans in service on fast speed. 6. Refer to FNP-2-S0P-12.1, CONTAINMENT AIR COOLING SYSTEM, for guidance on operating Q2P16MOV3024A (B, C, D), EMERG SW FROM 2A (B, C, D) CTMT CLR. 7. Refer to the Technical Specifications, LCO 3.6.5, for checking Containment average air temperature and LCO Requirements.

References:

A-207100, Sh. 98; B-205968; D-205010, Sh. 1; Technical Specifications; PCN B-86-2-3923; MDFD 88-1907; PCN B91-2-7619; PCN B91-2-7574, U-280362 (Vendor manual for N2T12TRSH3188) Page 2 of2 Version 31.0 07/02/09 06:48:57 FNP-2-S0P-12.1 NOTE: 2A CTMT CLR fan fast speed breaker EA-10-2 is interlocked with tie breaker EA08-2 to prevent starting if the emergency section of 2A 600V LC is aligned to 2D 600V LC (EA08-2 open). DCP B87 4592 4.1.6 Start 2A, 2B, 2C, and 2D containment coolers in FAST (SLOW) speed.

2A containment cooler
2B containment cooler
2C containment cooler
2D containment cooler 4.1.7 Check CTMT CLR 2A(2B,2C,2D)

DISCH 3186A(B,C,D) OPEN light illuminated.

CTMT CLR2A DISCH 3186A
CTMT CLR 2B DISCH 3186B
CTMT CLR 2C DISCH 3186C
CTMT CLR 2D DISCH 3186D 4.1.8 Place 2A, 2B, 2C, and 2D CTMT DOME RECIRC FANS in HIGH (LOW) speed.
2A CTMT DOME RECIRC FAN
2B CTMT DOME RECIRC FAN
2C CTMT DOME RECIRC FAN
2D CTMT DOME RECIRC FAN 4.1.9 Operate the containment dome recirculation fans and containment coolers as necessary to maintain containment temperature below "i20°F. (See section 4.7 for shifting containment cooler fan speeds.) ..----4.1.10 Open 2A and 2B RX CA V CLG DMPRS
2A RX CA V CLG DMPR Q2E12HV3999A
2B RX CA V CLG DMPR Q2E12HV3999B 4 Version 24.0 07/0210906:48:57 FNP-2-S0P-12.1
AUTION! ;, .. :tl1MERG.SW FROM 2A(2B,2C,2D)

CTMT CLRvalves Q2P16MOV3024A(B,C,D) [Q2P16V043A(B,C,D)1arenormally maintained maybe opened fortemperatnre control. However,operation witli these valves open s.hould be minimizedtorednce the potential.for long degradation to the containment coolers from the higher fl()w rates . . Shift Manager should . be consulted prior to opening these valves. 4.1.11 IF necessary to maintain containment temperature below 120°F, THEN open and caution tag EMERG SW FROM 2A(2B,2C,2D) CTMT CLR valves, as desired. 4.1.12

EMERG SW FROM 2A CTMT CLR VALVES Q2P16MOV3024A
EMERG SW FROM 2B CTMT CLR VALVES Q2P16MOV3024B
EMERG SW FROM 2C CTMT CLR VALVES Q2P16MOV3024C
EMERG SW FROM 2D CTMT CLR VALVES Q2P16MOV3024D WHEN additional service water flow through the containment coolers is no longer necessary for temperature control, THEN remove caution tag(s) and close EMERG SW FROM 2A(2B,2C,2D)

CTMT CLR valves:

EMERG SW FROM 2A CTMT CLR valves Q2P16MOV3024A
EMERG SW FROM 2B CTMT CLR valves Q2P16MOV3024B
EMERG SW FROM 2C CTMT CLR valves Q2P16MOV3024C
EMERG SW FROM 2D CTMT CLR valves Q2P16MOV3024D 5 Version 24.0

23. 022AK3.06 001/NEW/RO/MEM 3.213.3/N/N/4/CVRIY Unit 1 Reactor has just tripped, and the following conditions occurred:

All three RCPs have just tripped.
All Charging has been lost. Which one of the following correctly states the reason for maintaining CCW cooling flow to the Thermal Barrier HX in this condition?

Maintaining CCW cooling flow to the Thermal Barrier HX will prevent the RCP (1) from starting to degrade due to overheating in as early as (2) (1 ) (2) A. #1 seal 2 minutes #1 seal 13 minutes C. lower radial bearing 2 minutes D. lower radial bearing 13 minutes A -Incorrect. Incorrect, since the seal area takes time to void of the cooler water prior to allowing the hotter RCS water to the seal area. The WOG background document for ECP-O.O gives 13 minutes for the time to degrade the RCP #1 seal after losing CCW to the thermal barrier and seal injection. The #1 seal degrading is the correct concern and reason, but the 2 minutes is incorrect. Plausible, because with loss of CCW to the motor oil coolers the upper and lower MOTOR bearing (but not the radial bearing) can overheat in a maximum of 2 minutes per UOP-1.1 Step 3.10 (P & L). And, the lower RADIAL bearing would heat up in the event that both Seal Injection and CCW Thermal Barrier cooling were lost, but that is not the limiting concern or reason. B -Correct. The WOG background document for ECP-O.O gives 13 minutes for the time to degrade the RCP #1 seal after losing CCW to the thermal barrier and seal injection. C -Incorrect. The bearing is incorrect, since it is the #1 seal that is the limiting concern. Plausible, since the bearing will heat up in the condition given without CCW cooling to the Thermal Bearing, but the #1 seal is the limiting condition. The time is incorrect also, since the 13 minutes is given in the WOG background document for ECP-O.O. Plausible, since the 2 minutes would apply to an overheat RCP manual trip criteria in 2 minutes or less for a lower motor bearing if motor oil CCW cooling is lost (but not for lower RADIAL bearing). D -Incorrect. The first part is incorrect (see C). The second part is correct (see B). WOG Background Document FNP-O-ECB-O.O, LOSS OF ALL AC POWER, Plant Specific Background Information (pgs39 & 40 ci 88). Page: 60 cl200 1211412009 loolating the RCPthermai ba-rier CCW raurn outside containment ioolation valve prepares the plant for recovery whi I e protecti ng the CCW system from stearn formati on due to RCP thermal ba-ri er heati ng. Foil owi ng the loss of all ac power, hot reactor cool ant wi II grcduall y repl ace the normal I y cool seal injection water in the RCP seal a-ea. As the hot reactor coolant leaks up the shaft, the water in the thermal ba-rier will heat up and potentially form stearn in the thermal ba-rier and in the CCW lines cdj acent to the thermal ba-ri er. SubSffluent automati c start of the CCW pump woul d del iver CCW flow to the thermal ba-ri er, fl ushi ng the stearn into the CCW system. I f abnormal RCP seal leakage hcd developed in a pump, the abnormally high Iffi<age rate could exceed the cooling capacity of the CCW flow to that pump thermal ba-ri er and tend to generate more stearn in the RCP thermal ba-ri er CCW raurn lines. 1001 ati ng these lines prevents the potenti aI i ntroducti on of thi s stearn into the mai n porti on of the CCW system upon CCW pump start. Thi s keeps the ma n porti on of the CCW system avai I abl e for cool i ng *qui pment necessa-y for recoveri ng the pi ant when ac power is restored. KnOlNledge:

1. RCP seal integrity concernsfollowing loss of ac power (See Subsoction 2.1). 2. Analyses of RCPseaI performancefollowing a loss of all seal cooling esti mate that increased seal leakage may begin as early as 13 minutes due to seal degradation at high fluid temperatures.

It is important to establ ish suffi ci ent backpressure in the seal I ffi<off line by i 001 ati ng the seal raurn line before seal degrcdation occurs in order to limit RCP seal leakage. Thetirne of 13 minuteswas daermined in WCA P-1 0541 as thetime when " ... theIOlNer pump internals volume will be canpletely purged and the seal area water temperature will be approaching the 5500F reactor coolant temperature." FSD, CVCSlHHSl/ACCUMULATORlRMWS, A-181009 2.2.3.2 This capabil ity satisfies the seal water requi rement for the RCP No. I seal. A portion of the seal injection flow (nominally 5 gpm per pump) enters the RCS through the labyrinth seals and the thermal barrier. This in-leakage precludes leakage of reactor coolant through the No. I seal duri ng normal operati on. The remai nder of the seal i nj ecti on flow (nomi nally 3 gpm) flows up the pump shaft, cooling the pump lower bearing and the No. I seal. The 5-micron filtration requirement is based upon RCP minimum seal clearances. (Reference 6.2.44) FNP-1-SOP-1.1, Version 40.0 3.10 IF CCWfiow to the RCPmotor bearing oil coolers is lost, THEN pump operation may be conti nued unti I the motor upper or lower beari ng temperature reaches 195 0 F (approximately 2 minutes after cooling water flow stops). Page: 61 c:I 200 1211412009 Previous NRC exam history if any: 022AK3.06 022 Loss of Reactor Coolant Makeup AK3. KnCNVIedge of the reaSXlsfor thefollCMIing responses as they apply to the Los; of Reador Coolant Makeup: (CFR 41.5, 41.10/45.6/45.13) AK3.06 RCPthermai bcrrier cooling ................................ 3.2 3.3 Match justification: To correctly answer this question, knowledge is required of the reason for requiring RCP thermal barrier cooling during a loss of all RCS makeup (which would include a loss of seal injection). Objective:

2. RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the components associated with the Reactor Coolant System (RCS) to include the components found on Figure 1, Reactor Coolant System (OP&40301A02).

4. LABEL AND I LLUSTRA TE Reactor Coolant System (RCS) flow paths to include the components found on Figure 1, Reactor Coolant System (OPS-40301A05).

Page: 62 of 200 1211412009 06/27/0716:11:08 LOSS OF ALL AC POWER Plant Specific Background Information FNP-O-ECB-O.O Unit 1 ERP Step: 8 Section: Procedure Unit 2 ERP Step: 8 ERG Step No: 8 ERP StepText: Isolate RCP seals using ATTACHMENT

3. ERG StepText:

Dispatch Personnel To Locally Close Valves To Isolate RCP Seals And Place Valve Switches In CLOSED Position Purpose: To isolate the RCP seals Basis: This step groups three actions, with different purposes, aimed at isolating the RCP seals. The actions are grouped since all require an auxiliary operator, dispatched from the control room, to locally close containment isolation valves (the reference plant utilizes motor operated valves for the RCP seal return, RCP thermal barrier CCW return lines and RCP seal injection lines). This grouping assumes that the subject valves are located in the same penetration room area and that they are accessible. Concurrent with dispatching the auxiliary operator, the control room operator should place the valve switches for the motor operated valves in the closed position so that the valves remain closed when ac power is restored. Isolating the seal return line prevents seal leakage from filling the volume control tank (VCT) (via seal return relief valve outside containment) and subsequent transfer to other auxiliary building holdup tanks (via VCT relief valve) with the potential for radioactive release within the auxiliary building. Such a release, without auxiliary building ventilation available, could limit personnel access for local operations. Isolating the seal return line also enables pressure in the number 1 sealleakoff line to increase up to the relief valve setpoint of 150 psig. Maintaining a backpressure in the sealleakoff line of at least 150 psig enables development of high pressure in the number 1 sealleakoff cavity with a steady-state seal leakage rate established due to the self-limiting leakage characteristic of the number 1 seal. Under these conditions, with the number 1 seal functioning as expected and the number 2 seal remaining closed, the expected leakage flow rate is 21.1 gprn/pump. This is consistent with the state pressure distribution and seal leakage determined in the WCAP-10541 analysis and used in the latest RCP seal leakage PRA model in WCAP-15603. Isolating the RCP seal injection lines prepares the plant for recovery while protecting the RCPs from seal and shaft damage that may occur when a charging/SI pump is started as part of the recovery. With the RCP seal STEP DESCRIPTION TABLE FOR ECA-0.OStep8 injection lines isolated, a charging/SI pump can be started in the normal charging mode without concern for cold seal injection flow thermally shocking the RCPs. Seal injection can subsequently by established toJhe..RCP-consistenLwi1h.appropriate-plant.specific.prGGedw:es( Isolating the RCP thermal (barrier CCW return outside containment isolation valve prepares the plant for recovery while \ !rotectin g the CCW system from steam formation due to RCP thermal barrier heating. Following the loss of all ac power, hot reactor coolant will gradually replace the normally cool seal injection water in the RCP seal area. As the hot reactor coolant leaks up the shaft, the water in the thermal barrier will heat up and potentially form steam in the thermal barrier and in the CCW lines adjacent to the thermal barrier. Subsequent automatic start of the CCW pump would deliver CCW flow to the thermal barrier, flushing the steam into the CCW system. If abnormal RCP seal leakage had developed in a pump, the abnormally high leakage rate could exceed the cooling capacity of the CCW flow to that pump thermal barrier and tend to generate more steam in the RCP thermal barrier CCW return lines. Isolating these 39 of 88 Version: 1.0 06/27/0716:11:08 Knowledge:

References:

LOSS OF ALL AC POWER Plant Specific Background Information Section: Procedure FNP-O-ECB-O.O lines prevents the potential introduction of this steam into the main portion of system upon CCW pump start. This keeps the main portion of the CCW system for cooling equipment necessary for recovering the plant when ac power is restored. _ . RCP seal inte rit concerns followin loss of ac power (See Subsection 2.1). 2. Analyses of R seal per following a loss of all seal cooling es 1 a e seal leakage may begin as early as 13 minutes due to seal degradation at high'ffi:iid1eiiij?efatures. It 'lSTriijJ6rtant to establiSh sufficient backpressure in the line by isolating the seal return line before seal degradation occurs in order to limit RCP seal leakage. The time of 13 ,_minutes was determined in WCAP-1 0541 as the time when" ... the lower pump internals volume will be completely purged and the seal area water temperature will be the 550°F reactor coolant temperature." .). Time-critical actions of this step (for example, local seal return isolation) may be located earlier in the guideline if necessary to meet individual plant capabilities and requirements. Note that the Step Sequence Requirements allow interchangeability between Steps 6, 7 and 8 of this guideline. Justification of Differences: Placed actions in an Attachment to allow an extra operator to perform required actions outside of the control room without interfering with the flow of the procedure. 400f88 Version: 1.0 01/03/08 12:40:36 FNP-I-S0P-l.l 3.6 3.7 3.8 3.11 3.12 DO NOT attempt to start a RCP unless its oil lift pump has been delivering oil to the upper thrust shoes for at least two minutes. Observe the oil lift pumps indicating lights to verify correct oil pump motor operation and oil pressure. The oil lift pumps should run at least 1 minute after the RCP's are started. An interlock will prevent starting a RCP until 600 psig oil pressure is established. Shift Supervisor's approval must be obtained prior to removing any seal wires or changing the position of any throttle valves. RCP seal water injection flow of 6 gpm or CCW to the RCP thermal barrier must be continuously supplied when RCS temperature exceeds 150°F. Maintain RCP CCW and seal injection water supply temperature less than 105°F and 130°F respectively. IF CCW flow to the RCP motor bearing oil coolers is lost, THEN pump operation may be continued until the motor upper or lower bearing temperature reaches 195°F (approximately 2 minutes after cooling water flow stops). For RCP operations, a minimum pressure differential of200 psid must be maintained across RCP No. 1 seals. The following precautions apply in the case of a RCP # 1 seal failure. 3.12.1 3.12.2 DO NOT restart an RCP with an indicated No.1 seal failure. Refer to FNP-I-ARP-l.4, MAIN CONTROL BOARD ANNUNCIATOR PANEL "D", for guidance if No. 1 sealleakoffflow is abnormally low (Ann. DC 1) or abnormally high (Ann. DC2). 3.13 The No. I seal bypass valve should NOT be opened unless either the pump bearing temperature (seal inlet temperature) or the No. I sealleakofftemperature approaches its alarm level. The No. I seal bypass valve should then be opened only if all of the following conditions are met: 3.13.1 3.13.2 3.13.3 3.13.4 Reactor coolant system pressure is greater than 100 PSIG AND less than 1000 PSIG. No. I sealleakoffvalve is open. No. I sealleakoff flowrate is less than one GPM. Seal injection water flow rate to each pump is greater than six GPM. 3.14 For RCP operations, the required minimum back pressure of 15 psig on the RCP No.1 seals is ensured by maintaining a pressure of at least 18 psig in the VCT. Version 40.0 FNP Units 1 & 2 WESTINGHOUSE PROPRIETARY CLASS 2 CVCSIHHSIIACCUMULATORlRMWS A-181009 2.2.2.15 These requirements are to be satisfied assuming the nominal excess letdown flow rate (12,400 lbmlhr), normal RCP No. I seal leakage (3 gpm) from each RCP, 60 gpm recirculation flow from one charging pump, and normal VCT pressure. (Reference 6.2.1) The CVCS is required to makeup for shrinkage during a 100 degree F/hr cooldown of the RCS from hot zero power to 350°F. This is considered to be an original design basis function of the CVCS. 2.2.3 Seal Injection and Leakoff 2.2.3.1 2.2.3.2 2.2.3.3 The CVCS is required to cool excess letdown, RCP No. I seal leakage, and recirculation flow from at least one charging pump to 115°F prior to returning the flow to the charging pump suction. The cooling function is performed by the excess letdown heat exchanger and the seal water return heat exchanger. The 115°F temperature is based upon the RCP seal injection temperature limit of 130°F. (References 6.2.44, 6.2.7 and 6.2.8) The CVCS is required to provide a seal water injection flow rate adjustable over a normal range of 8 to 13 gpm to each RCP No. I seal. It is required that suspended solid particles larger than 5 microns in size be removed from the injection stream. This capability satisfies the seal water requirement for the RCP No. I seal. A portion of the seal injection flow (nominally 5 gpm per pump) enters the RCS through the labyrinth seals and the thermal barrier. This in-leakage precludes leakage of reactor coolant through the No. I seal during normal operation. The remainder of the seal injection flow (nominally 3 gpm) flows up the pump shaft, cooling the pump lower bearing and the No. I seal. The 5-micron filtration requirement is based upon RCP minimum seal clearances. (Reference 6.2.44) The CVCS is required to provide a means for cooling the RCP lower bearing under low RCS pressure conditions when the RCP No. I sealleakoff flow is less than 1.0 gpm. 2-12 Rev. 18

24. 026AA 1.01 001/NEW/RO/C/A 3.1/3.1/N/N/2ICVR/Y Unit 1 is at 100% power, and the following conditions occurred:

At 1000:

B Train CCW is on service, and aligned for split train operation.
A loss of 1 L 4160V Bus has occurred.

At 1015:

RCP motor bearing temperatures are: 1A: 172°F and rising 1 B: 165°F and rising 1 C: 19rF and rising At 1020:
RCP bearing temperatures are: 1A: 230°F and rising 1 B: 221°F and rising 1 C: 240°F and rising Which one of the following states: 1) the procedure(s) that must be entered, and 2) the EARLIEST time that a reactor trip is required based on RCP Bearing T emperatu res? (1 ) (2) A. AOP-10.0 ONLY 1015 B. AOP-10.0 ONLY 1020 Both AOP-9.0 AND 10.0 1015 D. Both AOP-9.0 AND 10.0 1020 Page: 63 of 200 1211412009 A -I ncorrect.

First part is incorrect since entry conditions are met for both AOP-1 0 & AOP-9.0. They would be done in parallel. AOP-10 would direct AOP-9 to be entered if it was not entered directly. AOP-1 0 ONLY is plausible, since the loss of SW is the initiating event and AOP-1 0 would deal with it. Second part is correct since the temperature for requiring a reactor trip on high RCP Motor bearing temperatures (195°F) per step 2 of AOP-9.0 have been exceeded. B -Incorrect. The first part is incorrect (see A). The second part is incorrect (see A). Plausible, since per AOP-4.1 , Abnormal Reactor Coolant Pump Seal Leakage, there is a RCP trip criteria for "CHECK RCP lower seal water bearing and seal water outlet temperatures stabilize less than 225°F". Confusion may exist between the two temperatures at which to trip the Reactor and RCP: 195 & 225. C -Correct. The AOP-1 0 entry is required per the loss of the B train SW Buss, and loss of B train SW, per the entry conditions. AOP-10 will also direct entry into AOP-9 further into the procedure, and the entry conditions for AOP-9, Loss of CCW, directs entry into AOP-9.0 for a loss of SW supplying an operating CCW train. Second part is correct since 1015 is the earliest that 195°F RCP bearing temperature is exceeded, and this is the requirement for a reactor trip per the continuing action STEP 2 and associated note in AOP-9.0. Note in AOP-9 also states that if flow is not adequate to maintain temperatures, trip the reactor. D -Incorrect. The first part is correct (see C). The second part is incorrect (see A). AOP-9.0 Ver. 22 AOP-1 0.0 Ver. 15 FNP-1-AOP-4.1, Abnormal Reactor Coolant Pump Seal Leakage, Version 5.0 _ 7 CH ECK RCP lower seal water becri n9 and seal water outlet temperatures stabi I i zes I ess than 225 0 F. FNP-1-AOP-9.0, Loss Of Canponent Cooling Water, Version 22.0

Check RCP motor becri n9 temperatures I ess than 195 0 F. Page: 64 of 200 1211412009 Previous NRC exam history if any: 026AA1.01 026 Loss of Component Cooling Water AA1. Ability to operate and lor monitor the following as they apply to the Loss of Component Cooling Water: (CFR 41.7/45.5/45.6)

AA 1.01 CCW temperGture indicaions ...................................... 3.1 3.1 Match justification: A loss of CCW is provided in the question, and rising temperature values are given. Knowledge is required to answer the question of how to operate as a result of the temperatures and monitor the temperatures (i.e which procedure(s) is(are) entered for this condition, and at which temperatures a reactor trip is required). Objective:

2. EVALUATE plant conditions and DETERMINE if entry into AOP-10.0, Loss of Service Water is required. (OPS-52520J02)

4. LIST AND DESCRI BE the sequence of major actions associated with AOP-10.0, Loss of Service Water. (OPS-52520J04).

5. EVALUATE plant conditions and DETERM I NE if any system components need to be operated while performing AOP-10.0, Loss of Service Water. (OPS-52520J06).

2. EVALUATE plant conditions and DETERMINE if entry into AOP-9.0, Loss of Component Cooling Water is required. (OPS-52520102)

6. EVALUATE plant conditions and DETERM I NE if any system components need to be operated while performing AOP-9.0, Loss of Component Cooling Water. (OPS-525201 06). Page: 65 of 200 1211412009 09/02/08 10:43:52 FNP-1-AOP-10.0 A. Purpose tt"J .. L\?i;L .. "L;! LOSS OF SERVICE WATER Version 14.0 This procedure provides actions for response to a loss of one or both trains of service water. This procedure is applicable at all times. B. Symptoms or Entry Conditions I. This procedure is entered when a loss of either train of service water is indicated by any of the following:

a. Actuation of SW PRESS A TRN LO annunciator AD4 or SW PRESS B TRN LO annunciator AD5 (60 psig) b. Actuation of SW PUMP TRIPPED annunciator AE4 c. Actuation of SW TO AUX BLDG HDR PRESS A OR B TRN LO annunciator AE5 (50 psig) d. Trip of any operating SW PUMP e. Rising temperatures on components supplied by service water f. Loss of power to one or both S W 4160 V busses 1 K or 1 L Page 1 of 16 11/25/087:42:14 FNP-1-AOP-9.0 A. Purpose LOSS OF COMPONENT COOLING WATER Version 22.0 This procedure provides actions for response to a loss of an operating component cooling water train. This procedure is applicable at all times. B. Symptoms or Entry Conditions I. This procedure is entered when a loss of component cooling water is indicated by any of the following:

a. Trip of any operating CCW PUMP b. Loss of SW supply to an operating CCW train Page 1 of 11 11125/087:42:14 FNP-1-AOP-9.0 LOSS OF COMPONENT COOLING WATER Version 22.0 NOTE: Action/Expected Response Response Not Obtained

Step 2 is a continuing action step.
IF RCP motor bearing temperatures exceed 195°F, THEN the ON SERVICE train is affected.
Adequate CCW flow means sufficient cooling is available to maintain acceptable temperatures.(i.e.

charging pumps, RHR cooling, SFP cooling, RCP's etc.)

Indications of pump cavitation are: Abnormal CCW flow oscillations or cavitation noise reported at the pump. 2 Check CCW system adequate for continued plant support. 2 Perform the following:

)\0 NOTE:

Check CCW flow adequate in affected train.
Check motor bearing temperatures less than 195°F.
Check CCW pump not cavitating.

Stop any cavitating CCWpump.

CCW Surge tank level being maintained at or above 13 inches. 2.1 IF the ON SERVICE train is affected, THEN perform the following:

2.1.1 IF the reactor is critical, THEN trip the reactor and perform, FNP-I-EEP-O, REACTOR TRIP OR SAFETY INJECTION, while continuing with this procedure. 2.1.2 Verify all Reactor Coolant pumps stopped. 2.1.3 IF in Mode 3 or 4, THEN perform FNP-I-AOP-4.0, LOSS OF REACTOR COOLANT FLOW while continuing with procedure. Indications of CCW pump cavitation will be abnormal CCW flow oscillations or cavitation noise reported at the pump. 2.2 IF evidence of CCW pump cavitation exists, THEN stop affected CCW pump. o Step 2 continued on next page Page 3 of 11 11125/087:41

37 FNP-1-AOP-4.1 Ii)",,,) "?:; S .2L i. ABNORMAL REACTOR COOLANT PUMP SEAL LEAKAGE Version 5.0 I I I Action/Expected Response Response Not Obtained -******************************************************************************************

CAUTION: To prevent potential seal damage, neither seal injection nor CCW cooling should be restored to a RCP which has lost both seal injection and CCW cooling. ****************************************************************************************** NOTE: Refer to the Integrated Plant Computer page lRCP, RCP Temperature Summary, for RCP seal water temperatures. 7 CHECK RCP lower seal water bearing 7 Perform the following: and seal water outlet temperatures stabiIlzes less than 225°F . . ----Monitor the following computer points for 7.1 Shutdown the affected reactor coolant the affected pump. pump as follows: [ ] TE 132 RCP A SEAL WATER OUTLET TEMP 7.1.1 Manually trip the reactor, AND go to [ ] TE0131 RCP A LOWER SEAL WATER FNP-I-EEP-O, REACTOR TRIP OR BRGTEMP SAFETY INJECTION. [ ] TEO 129 RCP B SEAL WATER OUTLET {CMT 0003908} TEMP [ ] TE 128 RCP B LOWER SEAL WATER BRGTEMP 7.1.2 WHEN the reactor is shutdown, [ ] TEO 126 RCP C SEAL WATER OUTLET THEN stop the affected RCP(s). TEMP {CMT 0003908} [ ] TE0125 RCP C LOWER SEAL WATER BRGTEMP 7.2 IF 1A OR IB RCP is secured, THEN close the pressurizer spray valve for the affected RCP. [ ] PK444C for lA RCP [ ] PK444D for IB RCP , Step 7 continued on next page Page Completed Page 8 of24

25. 026K1.01 001/NEW/RO/MEM 4.214.21N/N/3/CVRIY A Unit 1 Safety Injection is in progress due to a Large Break LOCA. Which one of the following describes the connection(s) between the RWST, A Train CS and ECCS pumps suction, and the operation of MOV-8827 A and MOV-8826A, CTMT SUMP TO 1A CS PUMP valves? A Train CS Pump, A Train HHSI Pump, and the A Train RHR Pump have (1) suction header(s) penetrating the RWST, and the CS Sump suction valves (2) automatically open on a LO-LO RWST condition.

(1 ) (2) A. separate will NOT B. one common will C. separate will 0:' one common will NOT A -. Incorrect. The first part is incorrect, but plausible since most of the safety related equipment has physical train separation for piping. The RWST is designed to minimize tank penetrations, and uses only one penetrations for suctions to all CS pumps, RHR pumps, and CVCS/HHSI pumps. The second part is correct. B -Incorrect. The first part is correct. The second part is incorrect, but plausible since this would be correct for the RHR sump suctions which have the auto function described. C -Incorrect. The first part is incorrect (See A). The Second part is incorrect (See B). o -Correct. The RWST is designed to minimize tank penetrations, and uses only one penetrations for suctions to all CS pumps, RHR pumps, and CVCS/HHSI pumps. The CS Sump suction valves do not have the auto open feature, but the RHR sump suctions do. Page: 66 of 200 1211412009 Previous NRC exam history if any: n/a 026K1.01 026 Contai nment Spray System K1 Knowledge of the physical connectionsandlor cause-effect relationships between the CSS and the (CFR: 41.2 to 41.9 / 45.7 to 45.8) K1.01 ECCS .......................................................... 4.24.2 Match justification: The only physical connection between the CS system and the ECCS system is at the RWST suction of the pumps, which is tested in the first part of each choice. The second part of the distractor contrasts the design of the CSS with the ECCS system sump suction valves to provide symmetry and three plausible but incorrect distractors. Objective: 1 LABEL AND I LLUSTRATE the Emergency Core Cooling System to include the components found on the following figures (OPS-40302C05): Page: 67 of 200

Figure 2, Accumulators
Figure 3, Refueling Water Storage Tank and Figure 4, Emergency Core Cooling System
The flow paths found on Figure 14, ECCS Injection Phase, Figure 15, ECCS Cold Leg Re:::irculation, Figure 16, ECCS Simultaneous Hot & Cold Leg Re:::irculation Normal, and Figure 17, ECCS Simultaneous Hot & Cold Leg Re:::irculation Alternate.

1211412009 Date: 1 0/21/2009 8 9 10 11 12 Time: 07:43:34 AM 13 0-3 501R 3/4-S1-2501R HYDRO TEST CONN. D-175038 SH.2 (E-ll) SIS. ACCUM TST. 110E388 SH.2 LOC. F-2 LOCAL SAMPLE 2" HCD-97 r--------------------<D 175039 SH, 7 (H 3l< E-6 . 110E388 SH.3 LOC. J-4 I > D-170118 )-, arl: V" I I YJ,7 12. 1 r; ,f I I ,"t' : SFPCS I I I ,,/I LV 110E400 LOC. J-8 :: tw'\.' HCO-96 I L .... ---..;.;;:,::.....:;"---_ilf_ "N:1 71'::-18-'- ---2-SH51R 1" HCD-96 l-SH51R 3/4-SI-2501R l-S1-2501R N1F16TA4145A IN.503-70 N1F16TE510 3/4" DR. f----tlM 3/4" V OV903 QV900 I. 3/4" v 1" QV904 N N I CD '-' '-' ;., / , ' 8 " QV022 1-8935 l-T58 r:i: cb c u :co ('>.j xg l-HCV-947.../"" l-RA58D l-S1-2501 R.../ 1 1/2" X 1" RED--./ 3/4" X 1/2" RED 1 3/4-X42D WHT DH HYDRO TEST PUMP ITEM: ALACSAPHD SAFTEY CLASS NNS N1E21P003-N OlUl(l) (3l/l(O r--- Ol 6 t3tJ I 0 I 01E13 12" HCB-152 D-175038 SH.3 E-12 SIS SPRAY PUMPS 110E388 SH.3 LOC. F-l 14" HCB-43 D-175038 SH.2 (F -11) 16" X 14" RED L I ,

SIS RHR PUMP 110E388 SH.2 LOC. D-2 L ---< D-175038 SH.2 (E -3) < RHR HX OUT 110E388 SH.2 LOC. E -11 6 (3 = 0"-u c:; NOTES: I i lA.

,'z. JM"l/'s Li> '-' eI 8 ","' ro =u .... :r:'j FOR SYSTEM OPERATION. '-' eI I r M , i!2 1i' 13 1 B. AIR AND ELECTRICAL POWER REMOVED. -{MY S : '-',lei I 6 15l:: NOT REQUIRED FOR 4" X 3" RED 1 I ,'7 8-C92 ASME SECTION III SEISMIC CATEGORY 1, I I I I 1-8926 RECIRCULATION SYSTEM WERE INSTALLED ... 1') I" -BUT HAVE BEEN DECLASSIFIED BASED ON TECH. I r-: :--'11--, 11C. PIPING AND COMPONENTS OF THE BIT/BIT 'T' C I r ? L,' 8" HCB-142 :r: SPEC. AMENDMENT

30. B I I I r --./ I IA 2. SAFETY RELIEF VALVE GAGGED DURING INITIAL 'lB r-.J-l><}-.J

... -!..-------.1-----..lL"L '\ " I S HYDROTEST AND SET TO 700 PSIG FOR 78 v , , II ,--"--------:><-----i rM'l/ NORMAL PLANT OPERATION. --' '------l : r M , '-j " I '-'lei 3. LOCATE TAPS 2'-0" APART. '2.. JM' I '-'lei NO. 1 L --C': 4, LOCATE CONNECTION ABOVE WATER LEVEL. Li' "eI I ,'-.A r M , 5, FLANGES FOR FLOW METERING ORIFICE I r-t. :-'Ir-, HIGH HEAD ,-"rL';, TO VERIFY FLOW DURING PRE-OPERATIONAL -'-I I SAFTEY INJECTION I 8-SH51R TESTING. )t -{MJ : (CHARGING) PUMPS

6. EQUIPMENT INSIDE BOX IN OTHERS SCOPE. T I "r / '--' L\ 7. 3 IN. 150 LB. R.F. STUDDING OUTLET r QV543 WITH 3 IN. BLIND FLANGE. I I

__ I 3/4" 0 8. 1 1/2 IN. 150 LB. R.F. STUDDING OUTLET. :----l><}-.J v ... r-!.-------.1-----..lL{

'\ " 1--------- --, 9. A 3/8" FLOW RESTRICTOR IS REQUIRED AS I: " , -j-,i----.;><-----, I NOTED ON LEGEND. > '2..JM' I r M , ,_/ ,,:1 I 10. ALL VALVE & LINE NUMBERS ON THIS DWG. .... L.' '-'eI I "lei NO.2 LM.r 7, I ARE PREFIXED BY Q1E21 EXCEPT WHERE ::g;::' : L-{':-_,:I-_,.

11.

WITH DWG. 0-175364 ; 0 :0'" FOR CVCS DETAILS SEE I

THRU 0 175372 .... << .;f -{M J (jf) DWG 110E386 SH 2 " :1 :-12. SEE owe. 0-175044 FOR WESTINGHOUSE

..... I "r / LM.T L'\ r.:l/s:r: I SYSTEM LEGEND. '" '" , LM.J ro I 13. FOR PIPING CLASS

SUMMARY

SHEETS SEE i5-I I II r --I I I SPEC. SS-1109-1. r-l><I_--v ... r---------.1----- ii"L, ____ .;><I_----,L-J 14. FOR VALVE INFORMATION SEE MASTER I / " \l VALVE LIST (SS-1102-39) I ,_/ I LM.TL" 15. TEMPORARY STRAINER IS PLACED IN SPOOL L ---.. NO. 3 I ,'7 PIECE DURING INffIAL FLUSHING OPERATIONS JM' I : LM . .r -;;, STRAINER MUST BE REMOVED BEFORE PLANT '-' eI I FOR CVCS DETAILS SEE STARTUP. I (jf)DWG 110E386 SH. 2 r M ,--} r-J 16. ADJUST VALVES IN FIELD DURING PRE-OPER-: : '-' eI L I , ATIONAL TESTING TO BALANCE INJECTION FLOW ,J, c ,0, 22. FLOW LIMITING ORIFICE INSTALLED TO ACHIEVE 4 00 PUMP RUN OUT, THEN LOCK INTO PROPER INJECTION FLOW. 'l' 17. DRAIN ALL DRAIN CONNECTIONS TO LOCAL L, I 23, LINE HCC-195 HAS BEEN DOWN GRADED FROM oS EQUIPMENT DRAINS. \/ ASME CLASS 3 SEISMIC CATEGORY 1 TO NNS "' g L,': SEISMIC CATAGORY II BY FSAR AMENDMENT 73 '" 'I :C N 18. REFER I I VALVE AND LINE NUMBERS HAVE NOT CHANGED. :i: "':c 19. DELETED. oJ FLAG (6) DESIGNATION NNS CLASIFICATlON, u r-'" 20 MULTIPLE LOOP PWR SUPPLY (CABINET MTD) I ---24. U262852-BOP PROCESS INSTRUMENT AND CONTROLS, "' g 2 . NOT SHOWN .

TRAIN A DWG. 7408023, TRAIN B DWG. 7408045. :I: '" t3 . 25. TE'S, AS NOTED ARE CLAMPED ON RTD'S FOR THERMAL I U 0 21. M(NUOLTTIPSLHEOWLONO.)P PWR SUPPLY CONTROL BOARD I-,;;-J I f'VI"'1 U(\lIITTf'oOlt..lf'

'" , '? (;;: B c D E F G H Title: C:\Reference Disk\Exam Reference Disk\Drawings\D175038-0001.cal FNP Units 1 & 2 3.3.3.2 3.3.3.3 3.3.3.4 RESIDUAL HEAT REMOVAL A-181002 The tank is designed according to the requirements of the ASME Boiler and Pressure Vessel Code Section III, Class 2. (Reference 6.1.29) Level Indication complies with the requirements of IEEE-279-1971, Criteria for Protection Systems for Nuclear Power Generating Stations. (Reference 6.1.33) Level Indication complies with the requirements of IEEE-308-1969, Criteria for Class IE Power Systems for Nuclear Power Generating Stations. (Reference 6.1.33) 3.3.4 I&C Requirements 3.3.4.1 3.3.4.2 The R WST must be provided with 2 level transmitters. One channel provides high, Technical Specification minimum, low, and low-low level alarms. The other channel provides low and low-low alarms. These channels provide the following alarm functions (References 6.2.14, 6.4.2,6.7.79 and 6.7.80): a. High Level Alarm -This alarm alerts the operator to a high level which could lead to tank overflow.

b. Technical Specification Minimum Level Alarm -Indicates that the RWST level is less than that required by the Technical Specifications during normal operations.

c. Lo Level Alarm -Alerts the operator that the ECCS pumps should be manually aligned for recirculation phase operation.

d. Lo-Lo Level Alarm -Alerts the operator that the CS pumps should be manually aligned for the recirculation phase. If an SI signal is present, this alarm will automatically open RHRlLHSI pump sump suction valves 8811A & Band 8812A & B. (References 6.7.43 [Section 6.3.2] and 6.7.42 [Section 6.3.2.4])

One of the two redundant level indicators in the control room must be operable for Post Accident Monitoring. (References 6.1.18, 6.7.15 and 6.7.16) 3-10 Rev.25 I

26. 027AKl.02 OOlINEW/RO/MEM 2.8/3.1/N/N/3/CVR/VER 5 EDITORIAL Unit 2 is at 50% power, and PT-444, PRZR PRESS, pressure transmitter has failed to the 2230 psig position.

Which one of the following describes the effects on PK-444A, PRZR PRESS REFERENCE controller, and the pressurizer liquid density due to this malfunction? PK-444A controller demand goes (1) and the density of the pressurizer liquid goes (2) (1 ) (2) A. down up B. down down C. up up up down A -Incorrect. The first part is incorrect (see D). Plausible, since if the PT had failed 6 psig higher (above 2235 psig), the proportional integral controller would integrate the error signal down until the PORV 444B opened and the sprays opened. Also, the spray valve controllers are controlled by the "master" controller and when the pressure must be increased, the demand goes down. Confusion could exist as which controller function is being described. The second part is incorrect. Plausible, since the spray valve controllers are controlled by the "master" controller and when their demand goes up pressure goes down and the liquid density goes up. Also, steam space density does go up in this condition, and the liquid specific volume goes up (and specific volume, not density, is the value given in the steam table for the property of the liquid). B -Incorrect. The first part is incorrect (see A). The second part is correct (see D). C -Incorrect. The first part is correct (see D). The second part is incorrect (see A). 0-Correct. The Proportional/Integral PRZR PRESS controller senses a low pressure and the demand starts integrating higher and higher. This first causes the spray valves to close and the proportional heaters increase output. Then, the backup heaters energize. The pressurizer liquid heats up and expands (density goes down) due to the increased heat input into the pressurizer liquid. The integral part of the controller continues to add to the error signal and PORV-445A opens due to actual pressure increasing to 2235 on PT 445. The pressure cycles around the setpoint of the PORV at 2235 psig with a higher pressurizer liquid temperature. Page: 68 of 200 12/14/2009 Previous NRC exam history if any: 027AK1.02 027 Pressurizer Pressure Control System Malfunction AKI. Knowledge ofthe operational implications of the following concepts as they apply to Pressurizer Pressure Control Malfunctions: (CFR 41.8 1 41.10 1 45.3) AKl.02 Expansion ofliquids as temperature increases ........................ 2.8 3.1 Match justification: To answer this question correctly, it must be recognized that for this particular malfunction of the PRZR Press control system, the pressurizer liquid heats up and expands due to pressurizer heaters energizing and sprays closing. The operational implications must also be understood in that this causes controller demand to go up (which would cause actual pressure go up until a PORV will lift: PORV-44SA). Objective:

2. RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the following components associated with the Pressurizer Pressure and Level Control System to include those items found on Figure 2, Pressurizer and Pressure Relief Tank, Figure 3, Pressurizer Pressure Protection and Control, and Figure 7, Pressurizer Level Protection and Control (OPS-5220IH02).

5. DEFINE AND EVALUATE the operational implications of abnormal plant or equipment conditions associated with the operation of the Pressurizer Pressure and Level Control System components and equipment to include the following (OPS-5220IH07):

Normal Control Methods
Abnormal and Emergency Control Methods
Automatic actuation including setpoint, if applicable
Protective Interlocks Actions needed to mitigate the consequence of the abnormality Page: 69 of 200 12/14/2009 PT-'/ t<<-;)dh r 1c..-i'lfl1 t/crv / f -c 1ftYt-t-Y Table 2: Saturated Steam: Pressure Table -------

.. -.----------- Abs Press. Temp Sat. Specific Volume Sat Sat. Enthalpy Sat. Sat. Entropy Sat. Abs Press. LbiSq In. Fahr Liquid Evap Vapor liquid Evap Vapor Liquid E .. ap Vapor lb/Sq In. p t vi V tg Vg h f h fg h g s j S fg Sg P n.08m 32,018 (},01602Z 3302.4 3302.4 0,0003 1075,5 W75.S 0.0000 2,1872 2.1872 0.08865 0,25 59.323 0,016032 1235.5 1235.5 27.382 1060.1 1087.4 0.0542 2,0425 2,0967 0,25 0,50 79,586 0.016071 641.5 641.5 47,623 1048,6 1096.3 0,0925 1.9446 2,0370 0.50 1.0 101.74 0,015135 333.59 333.60 59.73 1036.1 1105,8 0.1325 1.8455 1.9781 1,0 5.0 162,24 0.016407 73.515 73.532 130.20 1000,9 1l3Ll 0,2349 1.6094 1.8443 5.0 In.O 193,21 0,0!6592 38,404 38,420 161.26 982.1 1143.3 0,2836 1.5043 L7879 10,0 14.696 212,00 0,016719 26.782 25.799 180.17 970.3 1150.5 0,3121 1.4447 1.7568 14.696 15.0 213.03 0,016726 26,274 26.290 181.21 969.7 1150,9 0,3J37 1.4415 1.7552 15.0 20.0 227.96 0.016834 20.070 20,087 196,27 960.1 llS6,3 0.3358 1.3962 U320 20.0 30.0 250.34 0.017009 13.7266 13.7436 218,9 945.2 1164.1 0,3682 i.3313 1.6995 30,0 40.0 267.25 0,017151 10.4794 10.4965 236.1 933,6 lJ69.8 0.3921 1.2844 1.6765 40.0 50.0 281.02 0,017274 8.4957 8,5140 250.2 9:23.9 il74.1 DAli2 1.2474 1.6586 50.n SO.O 292.71 0,017383 7.1562 7.1736 262.2 915,4 !177.6 0,4273 1.2167 1.6440 60.0 70,0 302.93 0,01,482 6.1875 6,2050 272.7 907.8 ilBO,6 0.4411 !.l905 1.6316 10.0 BO,O 312,04 0,017573 5,4536 5A7ll 282.i 900,9 1183.1 0.4534 !.l675 1.6208 88,0 90.0 320,28 0.017659 4.8779 4,8953 290.7 894.6 lJB5.3 0.4643 U470 1.6113 90.0 100.0 327.82 0.017740 4.4133 4.4310 298.5 888,6 1187,2 0.4743 1.1284 1.6027 100.0 110.0 334.79 0,01782 4,0306 4.0484 305.8 883.1 1188,9 0.4834 Ul15 1.5950 110.0 120.0 341.27 0,01789 3.7097 3.7275 312.6 877.8 1190.4 0,4919 1.0960 1.5879 120,0 139.0 347.33 0.01796 3.4364 3.4544 319.0 872.8 1191.7 0.4998 1,0815 1.5813 130.0 140.0 353,04 0.01803 now 3.2190 325,0 868,0 1193.0 0,5071 1.0681 1.5752 140,0 150,0 358.43 0.01809 2.9958 3.0139 330,6 863.4 1194,1 0,5141 1.0554 1.5695 150.0 160.0 363.55 0,01815 2,8155 2.8336 336.1 859,Q 1195.1 0.5206 .1.0435 1.5641 160.0 110,0 368.42 0.01821 2.6556 2,6738 341.2 854,8 1196,0 0.5269 LOm 1.5591 170.0 180,n 373.08 O.Ql827 2,5129 2,5312 346.2 850.7 !l96,9 0,5328 1.0215 1.5543 1 BO.O 190.0 377.53 0,01833 2.3847 2.4030 350,9 846.7 1197,6 0,5384 1.0113 1.5498 190,0 20B,0 381.80 0.01839 2,2689 2.2873 355,5 842,8 1198.3 0,5438 L0016 1.5454 200,0 210.0 385,91 0.01844 2.16373 2.18217 359,9 839.1 1)99.0 0,5490 0,9923 1.5413 210,0 220.0 389,88 0.01850 2.06779 2.08629 364.2 835.4 1199,6 0,5540 0,9834 1.5374 220.0 230.0 393,70 0,01855 1.97991 1.99846 368.3 831.8 1200.1 0.5588 0.9748 1.5336 230.0 240.0 397.39 0.01860 1.89909 1.91769 372.3 828.4 1200.6 0,5634 0.9665 1.5299 240.0 250.0 400.97 0,01865 1.82452 1.84317 376.i 825.0 12GlJ 0,5679 0,9585 1.5264 250,0 260,0 404.44 0.01870 U5548 L77418 379,9 821.6 1201.5 0.5722 0,9508 1.5230 260,0 270.0 407.80 0,01875 1.69137 1.71013 383,6 818.3 1201.9 0.5764 0,9433 1.5197 270.0 280,0 4]1.07 Q.n1880 1.63169 1.55049 387,1 815.1 1202,3 0,5805 0,9361 1.5166 280,0 290.0 414.25 0,01885 1.57597 1.59482 390,6 812,0 1202,6 0.5844 0,9291 1.5J35 2BD.O 300.0 417.35 0,01889 1.52384 1.54274 394,0 808.9 1202.9 0,5882 0.9223 1.5105 300,0 350.0 431.73 0.01912 1.30642 L 409,8 794.2 J204.0 0,6059 0.8909 1.4968 350.0 444,60 0.01934 L14162 424.2 780A 1204,6 0,6217 0,8530 1.4847 400.0 '45&:28 .. "O:ij195if rm'zi4 lm62 ....... '4:11'J "ji:f7:S:'

r2l:"-,t;ir*********

0638ir* '(1:8318 lAng . 45tHr 50a.o 467,01 0.01975 0.90787 449.5 755.1 1204.7 0,6490 0.8148 1.4639 560.0 550.0 476,94 0,OJ994 0.82183 0,84177 460.9 743.3 1204,3 0,6611 0]936 1.4547 550,0 600,0 486,20 0,02013 0.74962 0.76975 471.7 732.0 1203.7 0,6723 0,7738 1.4461 SOO,O 650,D 494,89 0.02032 0,68811 0.70843 481.9 720,9 1202,8 0,6828 Q.7552 1.4381 650.0 100.0 503.08 0.02050 0.63505 0,55556 491.6 710.2 1201.8 0,6928 0.7377 1.4304 700.0 150,0 510,84 0,02069 0,58880 0,60949 500,9 699.8 1200.7 0.7022 0.7210 1.4232 150.0 800.0 518.21 0,02087 0,54809 0.568% 509.8 689,6 il99.4 0.7111 0.7051 1.4163 800,0 850.0 525.24 Q,02l0S O.SU97 0.53302 518.4 579,5 1198.0 0.7197 0.6899 L4096 850.0 900.0 531.95 0.02123 0.47968 0.50091 526.7 669,7 1196,4 0.7279 0.6753 1.4032 906.0 950.0 538,39 0.02141 0,45054 0,47205 534.7 660,0 1194.7 0,7358 0,6612 1.3970 950.0 1000.0 544.58 a,02i5S 0,42436 0.44595 542,6 650.4 1192.9 0)434 0.6476 1.3910 1000,0 1050.0 550.53 0,02177 0.40047 0.42224 550.1 640.9 119LO 0.7507 0,6344 1.3851 1050.0 nao.o 556.28 0,02195 0.37863 0.40058 557,5 631.5 !l89.1 0.7578 0.6216 1.3794 1100.0 1150.0 561.82 0,02214 0,35859 0.38073 564,8 622.2 1187,Q 0.7647 0,6091 L3738 1150,0 1200.0 557.19 0.02232 0.340.13 0,36245 571.9 613.0 1184,8 0.7714 0.5969 1.3683 1200,0 1250,0 572.38 0,02250 0.32306 0,34556 578.8 603.8 H82,6 0.7780 0,5850 1.3630 1250.0 1300,0 577.42 0,02269 0.30722 0,32991 585.6 594,6 1180,2 0.7843 0,5733 1.3577 130M 1350.0 58232 0.02288 0,29250 0.31537 592.3 585.4 1177.8 0,7906 0.5620 1.3525 1350,6 1400,0 587.Q7 0.02307 027871 Q,30178 598.8 576.5 H75,3 0.7966 0,5507 1.3474 1400.0 1450.0 591.70 0,02327 0.26584 0.289 it 605,3 567.4 1172.8 0,8026 0,5397 1.3423 1450,6 15QO,0 596.20 0,02346 0.25372 0.27719 611.7 5;;8.4 1170.1 0,8085 0,5288 1.3373 1500,0 1550.0 600.59 0,02366 0,24235 0.26601 618,0 549.4 1167.4 0.8142 0.5182 1.3324 1550.0 1600.6 604.87 Q,02387 0.23159 {),25545 624,2 540,3 1164,5 0.3199 0.5076 1.327.4 1600.0 1650,0 509.05 0,02407 0.22143 0,24551 630.4 531.3 1161.6 0,8254 0.497J 1.3225 165n,0 1700,0 613.13 Q,02428 0,21178 0,23607 636.5 522.2 !l58.6 0,8309 0.4867 .1.3176 1100.0 ',: 1750.0 ' 1150.0 617.12 0,02450 0.20263 0.22713 642,5 513.1 ]iSS,S 0,8363 0,4765 1.3128 1800.0 621.02 0.02472 0,19390 0.21861 648.5 503,8 1152.3 0,8417 0.4662 1.3079 1800.0 1850,D 624,83 {),02495 0.18558 0,21052 654.5 494.6 1l49,Q 0,8470 0.4561 1.3030 1650,0 1900.0 528.56 0.02517 0.17761 0.20278 660.4 485.2 1145,6 0,8522 0.4459 1.2981 1900.0 1950.0 632.22 Q,{)Z541 0.16999 0.19540 666,3 475.8 1142,{) 0,8574 {),4358 1.2931 1950.0 200U 535.80 0,02565 0.16266 0.18831 672.1 4662 1138.3 {),8625 0.4256 1.288i 2000.0 2100.0 642.76 0.02 " 0.14885 0.17501 683,8 446.7 1130,5 0,8727 0.4053 1.2780 210n,0 2200,0 649,45 0.13603 0.16272 426.7 1122,2 0,8828 0.3848 1.2676 2200,0 2308.0 655,89 0.12406 0.15133 707.2 406,0 liI3,2 0.8929 0,3640 1.2569 2300,0 2400.0 66Z.1l 0.11287 0.14076 719,0 384,8 li03.? 0,9031 0.3430 1.2460 24DIl.0 2500.0 668.11 0,02859 0.10209 0.13068 73U 361.6 1093.3 0,9139 0.3206 1.2345 2500.0 2600.0 673,91 0,02938 0.09172 0.12110 744,5 337.6 1082,0 0,9247 0,2977 1.2225 2600.0 2100.0 679.53 0,03029 0.08165 0.1lJ94 75i.3 312.3 1069.7 0.9356 0,2741 1.2097 2100.0 2800.0 684,96 0,03134 0,0717i O.103()5 770,7 285.1 1055,S 0.9468 0.2491 1.1958 2800.0 2900.6 69{),22 0.03262 0.06158 0.09420 785,1 254.7 1039.8 0,9588 0.2215 1,1803 2900.0 3000.0 695.33 0,03428 Q,050i3 Q.085 801.8 218A 1020,3 0,9728 0.1891 U619 3000.0 3100,0 700.28 0,03681 0,Q3i7l 0.07452 824,0 169,3 993.3 0.9914 0.1460 1.1373 3100.0 3200,0 705.08 0.04472 0,01191 0,05663 1l1!i.s 56.1 93L6 1.0351 0,0482 1.0832 3280,0 3208.2" 705,47 0.05078 0.00000 0.05078 906,0 0,0 906.0 1.0612 0.0000 1.0612 3208,2" f t:-j :j;; I\J 'Critlcal pressure :;::::;:;:::,;-;: __ "----"' .. ,,1)8:;-_ Abs Press LbiSq In. (Sat Temp) 1 (l0174) 5 062.2 4) 16 (193.21) 14.696 (212.00) 15 (213.03) 20 i227.961 25 (240.Q7) 30 (250.34) 35 (259.29) 46 (26725) 45 (27444) 59 (281.02) 55 (287.011 60 (292.71) 65 1297.98) 10 (302.93) 15 (307.61) Table 3. Superheated Steam Sat. Sat Temperature -Degrees fahrenheit Water Steam 200 250 300 350 400 450 500 600 100 800 900 1000 1100 1200 Sh 98.26 148.26 198.26 24826 298.26 348.26 398.26 49826 598.26 69826 99826 1098.26 0.01614 333.6 392.5 422.4 452.3 482.1 511.9 541.7 571.5 63U 690} 750.3 198.26 809.8 1483.8 2.3934 898.26 809.4 1534.9 24296 9290 988.5 69.73 1105.8 1150.2 1172.9 l!95.7 1218.7 1241.8 1265.1 1288.6 1336.1 1384.5 1433.7 0.1326 1.9781 2.0509 2.0841 2.1152 2.1445 2.1722 2.1985 2.2237 2.2708 2.3144 2.3551 1586.8 l639.7 2.4640 2.4969 Sh O.OI64J 73.53 130.20 113U 0.2349 1.8443 37.76 87.76 137.76 187.76 237.76 287.76 337.76 437.76 537.76 637.76 737.76 837.76 937.76 1037.76 78.14 84.2190.24 96.25 102.24 108.23 114.2] 126.15 138.08 150.01 161.94 173.86 185.78 197.70 1148.6 !l7L7 li94.8 1218.0 1241.3 1264.7 1288.2 1335.9 1384.3 1433.6 1483.7 1534.7 1586.7 1639.6 1.8716 1.9054 1.9369 1.9664 1.9943 20208 2.0460 2.0932 2.1369 2.1776 2.2159 2.2521 2.2866 2.3194 Sh 6.79 56.79 106.79 156.79 206.79 256.79 306.79 406.79 506.79 606.79 706.79 806.79 906.79 1006.79 0.01659 38.42 38.84 41.93 44.98 48.02 51.03 54.04 57.04 63.03 69.00 74.98 80.94 86.91 92.87 98.84 161.26 1143.3 1146.6 mO.2 1193.7 1217.1 1240.6 1264.1 1287.8 1335.5 1384.0 1433.4 1483.5 1534.6 1586.6 1639.5 0.2836 1.7879 L7928 1.8273 L8593 1.8892 1.9173 1.9439 1.%92 2.0166 2.0603 2.1011 2.1394 2.1757 2.2101 2.2430 Sh Sh Sh .ow 180.17 .3121 20.199 1150.5 1i,68 0.01673 26.290 181.21 1150.9 0.3137 i7552 , 0.01683 20081 196.27 1156.3 0.3358 L7320 Sh Sh Sh Sh Sh Sh 0.01593 16.301 208.52 mO.6 0.3535 1.7141 0.01701 13.744 218.93 1164.i 0.3582 1.6995 Ml108 11.896 228.03 !l6ll 0.3809 1.6872 0.01715 10.497 236.14 1169.8 0.3921 1.6765 0.01721 9.399 243.49 1172.l 0.4021 1.6671 0.01727 8.514 250.21 1I74.l OAll2 1.6586 Sh 0.01733 li87 256.43 illS.O 0.419& 1.6510 Sh v 0.01738 7.174 h 262.21 1177.6 0.4273 1.6440 Sh Sh 0.01743 6.653 267.63 1179.l 0.4344 1.6375 v 0.01748 6.205 h 272]4 1180.6 O.4411 1.6316 Sh 0.01753 5.814 277.56 118J.9 0.4474 1.6260 38.00 28.42 1168.8 1.7833 88.00 30.52 1192.6 1.8158 138.00 3260 l2lU 1.8459 lS8.00 34.67 1239.9 1.8743 238.00 3&.72 1263.6 1.9010 288.00 38.77 1287.4 1.9265 388.00 42.86 1335.2 1.9739 488.00 4693 1383.8 2.0177 588.GO 5l.0D 1433.2 2.0585 688.00 55.06 1483.4 2.0969 788.00 59.13 1534.5 2.1332 8go.00 63.19 1586.5 2.1676 988.00 51.25 1639.4 2.2005 36.97 86.97 136.97 i86.97 236.97 286.97 386.97 48697 586.97 27837 29.899 31.939 33.963 35.977 37.985 41.986 45.978 49.984 11687 1192.5 1215.2 1239.9 1263.6 1287.3 13352 1383.8 14312 !.7809 18134 1.8437 1.8720 1.8988 1.9242 1.9717 2.0155 2.0563 686.97 786.97 88697 53.946 57.926 6L905 1483.4 1534.5 1586.5 2.0946 21309 2.l653 985.97 55.882 1639.4 2.1982 22.04 72.04 122.04 172.04 20.788 22355 23.900 25428 ! 167.1 1191.4 1215.4 ,2392 1.7475 Ll80S l.8111 L8397 222.04 272.04 37204 472.04 572.04 672.04 772.04 872.04 972.04 26.945 28.457 31.466 34.465 37.458 40.447 43435 46.420 49.405 1263.0 1286.9 1334.9 1383.5 14329 1483.2 1534.3 1586.3 1639.3 L8666 1.8921 1.9397 1.9836 2.0244 2.0628 2.0991 2.!336 2.1665 9.93 59.93 109.93 159.93 209.93 259.93 359.93 459.93 559.93 659.93 759.93 859.93 95993 16.558 17.829 19.076 20.307 21.527 22.740 25.153 27.557 29.954 32.348 34.740 37130 39.518 1165.6 1190.2 1214.5 1238.5 1262.5 1286.4 1334.6 1383.3 1432.7 1483.0 1534.2 15862 1639.2 1.7212 L7547 1.7856 1.8145 1.8415 1.8672 1.9149 1.9588 1.9997 2.0381 2.0744 21089 2.1418 49.66 14.810 1189.0 1.7334 99.66 149.66 15.859 16.892 1213.6 1237.8 1.7647 1.7937 AO.71 90.71 140.71 12.654 13.562 14.453 1187.8 1212.7 1237.1 !.715;> 1.7468 1.7761 32.75 82.75 132.75 lL036 1l.838 12.624 !la6.6 121i.7 1236.4 1.6992 1.7312 1.7608 199.66 249.66 349.66 449.66 549.66 17.914 18.929 20.945 22.951 24.952 1261.9 1286.0 1334.2 i383.0 i432.5 1.8210 1.8467 1.8946 i.9386 1.9795 649.66 749.66 84966 949.66 26.949 28.943 30935 32.927 1482.8 1534.0 1586.1 1639.0 2.0179 20543 2.0888 2.!217 190]1 240.71 340]) 440.71 54Q.71 640.71 740)1 840)1 940.71 15.334 16.207 17.939 19.662 21.379 23.092 24.803 26.512 28.220 1261.3 1285.5 13319 1382.8 1432.3 1482.7 1533.9 1586.0 1638.9 1.8035 1.8294 1.8774 1.9214 1.9624 2.0009 2.0372 2.0717 2.1046 182.75 232.75 33275 43275 53215 632.75 732.75 832.75 93V5 13.398 14.165 15.685 17.195 18.699 20199 21697 23194 24.689 1260.8 1285.0 13336 1382.5 1432.1 1482.5 J5337 15858 16388 1.7883 1.8143 1.8624 1.9065 1.9476 1.9860 2.0224 2.0569 2.0899 25.56 75.56 125.56 175.56 225.56 425.56 525.56 625.56 725.56 825.56 925.56 9.777 10.497 11.201 11.892 12.577 13.932 15.276 16.614 17.950 19.282 20.613 21.943 1185.4 121004 1235.7 1260.2 1284.6 1333.3 1382.3 1431.9 1482.3 1533.6 1585} 1638} 1.6849 J.7173 U471 1.7748 1.8010 1.8492 1.8934 1.9345 1.9730 2.0093 2.0439 2.0768 18.98 68.98 118.98 168.98 2i8.98 31898 418.98 518.98 618.98 718.98 818.98 918.98 8.769 9.424 10.062 10.688 lL306 12.529 13741 14.947 16.150 17.350 18.549 19746 li84.i 1209.9 1234.9 1259.6 1284.1 1332.9 1382.0 1431.7 1482.2 1533.4 1585 S 1638.6 1.6720 1.7048 1.7349 L7628 1.7890 1.8374 1.8816 1.9227 1.9613 1.9977 2.0322 2.0652 12.93 7.945 1182.9 1.6601 729 7.257 1181.6 1.6492 2.02 6.675 1180.3 1.6390 62.93 8.546 1208.9 1.6933 57.29 7.815 1208.0 1.6829 112.93 162.93 m.93 312.93 412.93 512.93 612.93 712.93 812.93 912.93 9.130 9.702 10.267 11.381 12.485 13.583 14.677 15769 16.859 17.948 1234.2 1259.1 1283.6 1332.6 1381.8 1431.5 1482.0 1533.3 1585.5 1638.5 1.7237 1.7518 t7781 1.8266 1.8710 1.9121 1.9507 1.987 2.022 2.055 107.29 157.29 20729 30729 407.29 507.29 607.29 707.29 ... 807.29 8.354 8.881 9.400 10.425 11.438 J2.446 13.450 14.452 15.452 1233.5 1258.5 1283.2 1332.3 1381.5 1431.3 1481.8 l!i33.2 1585.3 1.7134 1.7417 1.7681 1.8168 1.8612 1.9024 1.9410 1.9774 2.0120 907.29 16.450 1638.4 2;045.0 52.02 102.02 152.02 202.02 302.02 402.02 502.02 602.02 702.02 7.195 7.697 8.186 8.667 9.615 10.552 11.484 12.412 13.337 1207.0 1232.7 1257.9 1282.7 1331.9 1381.3 143U 1481.6 1533.0 1.6731 Ll040 1.7324 Li590 1.8077 1.8522 1.8935 1.9321 1.%85 802.02 14.261 15852 2.0031 902.02 15.183 1638.3 2.0361 47.07 97.07 141.07 197.07 297.07 397.07 497.07 597.07 697.07 797.Q7 897.07 6.664 7.133 7.590 8.039 8.922 9.793 10.659 11.522 12.382 13.240 14.097 1481.5 1532.9 1585.1 1638.2 1.9238 1.9603 1.9949 2.0279 1206.0 1232.0 1257.3 1282.2 1331.6 1381.0 1430.9 1.6640 1.6951 1.7237 L7504 1.7993 1.8439 1.8852 4239 92.39 6.204 6.645 1205.0 1231.2 1.6554 1.6868 142.39 192.39 292.39 392.39 492.39 592.39 692.39 792.39 892.39 7.074 7.494 8.320 9.135 9.945 10.750 11.553 12.355 13.155 1256.7 1281.7 1331.3 1380.7 1430} 1481.3 1532.7 1585.0 1638.1 1.7156 1.7424 1.7915 l.8361 1.8774 1.9161 1.9526 L9872 2.0202 Sh "" superheat. F h = enthalpy, Btu per Ib s = entropy, Btu per R per Jb v = specific volume, cu It per Ib ::::

27. 027Kl.Ol OOllNEW/RO/MEM 3.4/3.7!N!N12/CVRIY Which one of the following correctly states how the Containment Spray System reduces radioactive iodine in the Containment atmosphere during a LOCA? To enhance absorption of Iodine from the Containment atmosphere, the Containment Spray System sprays water from the (1) at a pH of approximately (2) (1 ) (2) A. containment sump 4.5 B. RWST 7.5 containment sump 7.5 D. RWST 4.5 A -Incorrect.

First part correct, see C. second part -4.5 pH is incorrect. Plausible, since the Borated water from the RWST in the injection phase is a pH of approx. 4.5 due to the 2300-2500 ppm borated water. However, the recirc phase begins the spray of the sump water which has the dissolved Tri-Sodium Phosphate in it, and the pH of that water is higher at 7.5 to 10.5. B -Incorrect. The 7.5 is correct, but the RWST is acidic at a pH of 4.5 due to the high concentration of boric acid in. The low pH is not conducive to absorbing the iodine. The Iodine is absorbed during the recirc phase when the CS takes a suction on the Containment sump after the TSP has dissolved and raised the pH of the Spray water. Plausible, if confusion exists as to the need for the pH to be higher in order to absorb the Iodine out of the containment atmosphere. C -Correct. The TSP in the Containment Sump dissolves in the Containment sump water during the injection phase, and raises the pH from about 4.5 to a range of 7.5-10.5. During the Containment Spray recirc phase, this causes the iodine in the containment atmosphere to be absorbed in the spray water and convert to a non-volatile form. Then, it stays in the sump water, and does not leak out of containment via any ctmt atmosphere leakage paths. Even though some iodine would be absorbed by the mechanical action of the spray water in the containment atmosphere, the higher pH enhances the effect, and the retention of the iodine in the sump water is due to the higher pH. D -Incorrect. Both parts are incorrect (see A & C). Plausible, since the pH is correct for the RWST source, but this pH is not conducive to removing iodine from the containment atmosphere. Confusion may exist as to the exact mechanism of iodine removal by the CS system. TS 83.5.6 FSD A 181008, CS System Page: 70 of 200 12/1412009 2.0 SYSTEM FUNCTIONAL REQUIREMENTS The safety-related function of the CSS is to reduce the containment building pressure and temperature following a LOCA or high-energy line rupture and to reduce airborne fission products in the containment atmosphere following a LOCA. During the injection phase, the CSS pumps are aligned to take suction off the RWST. When the RWST reaches low-low level, the spray pumps operate in the recirculation mode from the containment sump. Operator action to perform realignment of the CSS pumps to sump recirculation must be completed within 130 seconds of reaching the RWST low-low level setpoint. Completion of this operator action in 130 seconds ensures sufficient volume remains in the RWST to ensure adequate pump NPSH is available and to prevent vortexing in the RWST (References 6.3.020,6.7.039). Trisodium phosphate (TSP) filled baskets in the recirculation area of containment provide iodine absorption and retention in the containment sump solution (References 6.2.001, 6.3.001, 6.7.001). As the RCS inventory combined with Eecs solution accumulates in the recirculation sump, the rising water level dissolves the TSP crystals in the baskets (References 6.7.033 and 6.7.034). The spray water is maintained at a pH level of approximately 4.5 during injection. During recirculation, a pH of approximately 7.5 enhances the absorption of the airborne fission product iodine, retains the iodine in the containment sump solution, and minimizes potential for chloride induced stress corrosion cracking (References 6.1.001, 6.2.001,6.3.001,6.3.017). The development of the iodine removal coefficient is a function of the characteristics of the CSS. The design value of the iodine removal coefficient is 10 hr -1. This coefficient is based on one CSS pump operating at a flow rate of2,200 gpm, and a spray fall height of 110 ft (References 6.3.018, 6.7.003). 3.1 CSS PUMPS 3.1.1 Basic Functions Post-LOCA, the CSS pumps shall deliver borated water from the RWST during the injection mode, water from the containment sump and trisodium phosphate from the TSP baskets during the recirculation mode, to the containment spray ring headers (References 6.2.001, 6.7.033, 6.7.034). Page: 71 of 200 12/1412009 Previous NRC exam history if any: Wrote a new question and intentionally stayed away from the 2008 nrc exam question on k/a 027G2.1.27, CS&COOL-40302D02 17 to prevent going over the limit of 4 RO questions from the previous 2 NRC exams. 027K1.01 027 Containment Iodine Removal System Kl Knowledge of the physical connections and/or cause-effect relationships between the CIRS and the following systems: (CFR: 41.2 to 41.9 1 45.7 to 45.8) K1.01 CSS ........................................................... 3.4* 3.7* Match justification: To answer this question correctly, the physical connections to the Iodine Removal and the CSS (only connected during the CS recirc phase taking a suction from the Sump instead of the RWST) , and the knowledge of the TSP (iodine removal) cause-effect on the CSS of adjusting the pH FROM 4.5 TO 7.5 or greater is required. Objective: 1 LABEL AND ILLUSTRATE the Emergency Core Cooling System to include the components found on the following figures (OPS-40302C05):

Figure 2, Accumulators
Figure 3, Refueling Water Storage Tank and Figure 4, Emergency Core Cooling System
The flow paths found on Figure 14, ECCS Injection Phase, Figure 15, ECCS Cold Leg Recirculation, Figure 16, ECCS Simultaneous Hot & Cold Leg Recirculation Normal, and Figure 17, ECCS Simultaneous Hot & Cold Leg Recirculation Alternate.

2 LABEL AND ILLUSTRATE the Containment Spray and Cooling System flow paths, to include the components found on Figure 2, Containment Cooling System, Figure 3, Containment Spray System and Figure 4, Service Water to Containment Coolers (OPS-40302D05).

2. RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the components associated with the Containment Spray and Cooling System to include the components found on Figure 2, Containment Cooling System, Figure 3, Containment Spray System and Figure 4, Service Water to Containment Coolers and the following (OPS-40302D02):

Containment Cooler Service Water Inlet Isolation Valves (MOV-3019A, B, C, and D)
Trisodium Phosphate Baskets Page: 72 of200 12/14/2009 ECCS Recirculation Fluid pH Control System B 3.5.6 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) B 3.5.6 ECCS Recirculation Fluid pH Control System BASES BACKGROUND The Recirculation Fluid pH Control System is a passive system designed to raise the long term pH of the solution in the containment sump following a Design Basis Accident (DBA). The Recirculation Fluid pH Control System consists of three storage baskets containing trisodium phosphate (TSP) as Na 3 P0 4
12H 20
Y-tNaOH. An equivalent amount of trisodium phosphate compound with a different chemical formula may be used. When equivalent compounds are used, the allowable weights/volumes may be different; however, the equivalent amount of trisodium phosphate comp' raise the pH of the recirculating solution into the range 0 7.5 to 10.5. In the event of a loss of coolant accident (LOCA), the SP contained in the storage baskets will be dissolved in the Reactor Coolant S stem RCS (a"ildFefueling Water Storage Tank (RWS Inventories lost through tJ:ie'"pTpe break. The resulting increase in the ecirculation so u ion H into the range of 7.5 to 10.5 assures that iodine is retalne In solution and that chloride induced stress corrosion on mechanical systems and components is minimized (Ref. 1). The Recirculation Fluid pH Control System performs no function during normal plant operation.

Radioiodine in its various forms is the fission product of primary concern in the evaluation of a DBA. Fuel damage following a DBA will cause iodine to be released into the reactor coolant and containment atmosphere. Iodine released to the containment atmosphere is r,,-s::'/A' 'i-:; '<""--f;; absorbed by the containment spray and washed into the containment -¥ \ i sump. Since the ECCS water is borated for reactivity control, the I7vJ 7') \.J recircuiation solu .. containment sump will initiall be acidic 0-L with a pH of approximately 4.5. In a low p (acidic) solution, some of () H::.J/ ... 7 the dissolved iodine will t3eC0nverted to a volatile form and evolve out lG '}. of solution into the containment atmosphere. In order to reduce the ;:yrp. _A potential for elemental iodine evolution, the ECCS recirculation solution v-lP" is adjusted (buffered) to achieve a long term alkaline pH of no fc ,than An alkaline pH promotes iodine hydrolysis, in which iodine is '1.-,.. converted to nonvolatile forms. In addition to ensuring iodine is retained in solution, an alkaline recirculation solution will minimize chloride induced stress corrosion cracking of austenitic stainless steel (continued) Farley Units 1 and 2 B 3.5.6-1 Revision 0

28. 029EAl.06 OOllMOD/ROICIA 3.2/3.11N1N12ICVRIY Unit 2 has experienced an Anticipated Transient Without Trip (ATWT) and the following plant conditions occurred:

Charging Flow is 68 gpm.
2A Boric Acid Transfer Pump is tagged out.
Safety Injection has NOT actuated at this time.
lAW FRP-S.1, Response to Nuclear Power Generation

-ATWT, the UO is establishing Emergency Boration.

2B Boric Acid Transfer Pump tripped when it was started. Which one of the following states: 1) the required actions to establish an emergency boration flow path, and 2) the MINIMUM required action for FK-122, CHG FLOW controller, lAW FRP-S.1? A. 1) Open V185, MAN EMERG BORATION valve, AND open FCV-113A, BORIC ACID TO BLENDER valve. 2) Place FK-122 in MAN ONLY. 1) Open LCV-115B and 0, RWST TO CHG PUMP valves, AND close LCV-115C and E, VCT OUTLET ISO valves. 2) Place FK-122 in MAN AND raise demand. C. 1) Open V185, MAN EMERG BORATION valve, AND open FCV-113A, BORIC ACID TO BLENDER valve. 2) Place FK-122 in MAN AND raise demand. D. 1) Open LCV-115B and 0, RWST TO CHG PUMP valves, AND close LCV-115C and E, VCT OUTLET ISO valves. 2) Place FK-122 in MAN ONLY. Page: 73 of 200 12/14/2009 A -Incorrect.

The first part is incorrect, since the Manual Emergency Borate flowpath will not work in this situation. Neither BAT pump is available, and at least one is required to use either the normal emergency or manual emergency borate flowpath. Plausible, since in other situations with a loss of the normal emergency borate flowpath, the MANUAL emergency borate flowpath would be used per FRP-S.1 Step 4.3 RNO. The second part is incorrect due to the charging flow being less than required (92 gpm) with the RWST boration flowpath aligned, and the charging will have to be increased by adjusting FK-122 in the raise direction per FRP-S.1, Step 4.6 & 4.7.3. Plausible, since the charging flow is greater than required for the normal emergency or manual emergency borate flowpath (40 gpm) per FRP-S.1 , step 4.6. B -Correct. Per FRP-S.1 Steps 4.2.1 RNO, the RWST boration flow path will be aligned due to the inability to start either BAT pump. The flow from the RWST to the RCS is required to be > 92 gpm per step 4.6, thus the charging demand must be raised in manual. C -Incorrect. The first part is incorrect (see A). The second part is correct (see B). 0-Incorrect. The first part is correct (see B). The second part is incorrect (see A). FRP-S.1 Revision 25 Previous NRC exam history if any: 029EA1.06 029 Anticipated Transient Without Scram (A TWS) EA1 Ability to operate and monitor the following as they apply to a ATWS: (CFR 41.7 145.5/45.6) EA1.06 Operating switches for normal charging header isolation valves .......... 3.2* 3.1 Match justification: To answer this question the applicant must know what to do with the operating switch for normal charging header isolation valve (Operate & monitor: FCV-122 Controlled by FK-122 controller), in response to the flow indication on FI-122 in the given situation during an ATWS. Objective:

6. EVALUATE plant conditions and DETERMINE if any system components need to be operated while performing (1) FRP-S.l, Response to Nuclear Power Generation/ATWT; (2) FRP-S.2, Response to Loss of Core Shutdown. (OPS-52533A06)

Page: 74 of200 12/14/2009 \\,) j\.i i 1 " FNP-1-FRP-S.1 RESPONSE TO NUCLEAR POWER GENERATION/ATWT Revision 25 Step n 3 Action/Expected Response Verify AFW pumps -RUNNING. 3.1 MDAFWPs -RUNNING [] 1A amps > 0 [] 1B amps > 0 3.2 TDAFWP -RUNNING IF NECESSARY

TDAFWP STM SUPP FROM 1B (IC) SG [] MLB-4 1-3 lit [] MLB-4 2-3 lit [] MLB-4 3-3 lit
TDAFWP SPEED [] SI 3411A > 3900 rpm
TDAFWP SPEED CONT [] SIC 3405 at 100% Response NOT Obtained NOTE:
2500 gallons of emergency boration is required for each control rod not fully inserted, up to a maximum of 17,309 gallons. 4
rCA] Emergency boration should continue until an adequate shutdown margin is established.

Initiate Emergency Boration of the ReS. 4.1 Verify at least one CHG PUMP -RUNNING. Step 4 continued on next page. _Page Completed Page 5 of 19 FNP-1-FRP-S.1 RESPONSE TO NUCLEAR POWER GENERATION/ATWT Revision 25 Step n Action/Expected Response 4.2 Start a boric acid transfer pump. ;J BATP lit? [] 1A .---;v [] 4.3 Align normal emergency boration. EMERG BORATE TO CRG PUMP SUCT [] Q1E21MOV8104 open Response NOT Obtained 4.2 Perform the following. 4.2.1 Align charging pump suction to RWST. RWST TO CRG PUMP [ ] Q1E2ILCV1l5 B open [] Q1E2ILCV1l5D open VCT OUTLET ISO [ ] Q1E21LCV1l5C closed [] Q1E21LCV1l5E 4.2.2 Perform the following.

Align charging pump suction to RWST. RWST TO CRG PUMP [] Q1E21LCV1l5B open [] Q1E21LCV1l5D open VCT OUTLET ISO [ ] Q1E21LCV1l5C closed [ ] Q1E21LCV1l5E closed
Align manual emergency boration flow path. BORIC ACID TO BLENDER [] Q1E21FCVl13A open MAN EMERG BORATION [] Q1E21V185 open (100 ft. AUX BLDG rad-side chemical mixing tank area) Step 4 continued on next page. _Page Completed Page 6 of 19

"",.!.1; 1'1 FNP-1-FRP-S.1 RESPONSE TO NUCLEAR POWER GENERATION/ATWT Revision 25 Step n Action/Expected Response 4.4 Establish adequate letdown. 4.4.1 Verify 45 GPM letdown orifice -IN SERVICE. LTDN ORIF ISO 45 GPM [] Q1E21HV8149A open 4.4.2 Verify one 60 GPM letdown orifice -IN SERVICE. LTDN ORIF ISO 60 GPM [] Q1E21HV8149B open [] QlE21HV8149C open 4.5 Check pressurizer pressure LESS THAN 2335 psig. 4.6 Establish adequate charging flow .

IF boration is from acid storage tank, fJ Response NOT Obtained 4.5 Verify PRZR PORVs and PRZR PORV ISOs -OPEN. IF NOT, THEN open PRZR PORVs and PORV ISOs as necessary until pressurizer pressure less than 2135 psig.

verify charging flow -I GREATER THAN _40 GPM .. ( It-q--{) l. j\.. J. OR

IF boration is from RWST, THEN verify charging GREATER THAN 92 GPM. -----------------------

Step 4 continued on next page. _Page Completed Page 7 of 19 !.'.; fi. .I. FNP-1-FRP-S.1 RESPONSE TO NUCLEAR POWER GENERATION/ATWT Revision 25 Step n Action/Expected Response 4.7 Verify emergency boration flow adequate. 4.7.1 IF normal emergency boration flow path aligned, THEN check emergency boration flow greater than 30 GPM. EMERG BORATE [] FI 110 4.7.2 IF manual emergency boration flow path aligned, THEN check boric acid flow greater than 30 GPM. MAKEUP FLOW TO CHG/VCT [] BA FI 113 4.7.3 IF boration is from the RWST, THEN verify charging flow -GREATER THAN 92 GPM. Page Completed Page 8 of 19 Response NOT Obtained

1. 029EAl.05 OOIlI/l/ATWT

-BIT OUTLET SW/CIA -3.7/MODIFIEDIRlNRC RO/TNT 1 RLM 029EA1.0 The following plant conditions exist: -Unit 2 has experienced an Anticipated Transient Without Trip (ATWT) and has implemented FRP-19211, Response to Nuclear Power Generation ATWT. -A Charging Pump is running. -Boric Acid Transfer Pump # 1 is tagged out. -Boric Acid Transfer Pump # 2 trips on start. -SI has NOT actuated at this time. -The SS has directed the RO to establish Emergency Boration in accordance with SOP-13009, "CVCS Reactor Makeup Control System". Which ONE of the following actions would establish a CORRECT emergency boration flow path in accordance with the SOP? (Assume 12 gpm seal return flow) A. 1) Open HV-8104 EMERGENCY BORATE Valve. 2) Adjust charging flow controller FIC-0121 to obtain> 42 gpm flow through the Normal Charging flow path. B. 1) Open LV-0112D and LV-0112E RWST TO CHARGING PUMP SUCT valves. 2) Adjust charging flow controller FIC-0121 to obtain> 42 gpm flow through the Normal Charging flow path. C. 1) Open FV-110A BA to Blender and FV-110B BLENDER OUTLET TO CHARGING PUMPS SUCT. 2) Adjust charging flow controller FIC-0121 to obtain> 100 gpm flow through the Normal Charging Path. 1) Open LV-0112D and LV-0112E RWST TO CHARGING PUMP SUCT valves and HV-8801A and HV-8801B BIT DISCHARGE ISOLATION valves. Page: lof3 2) Verify BIT flow (FI-0917 A), plus total seal injection flow, minus total seal return flow is > 100 gpm. 10/26/2009 KIA 029 Anticipated Transient Without Scram (ATWS): EA1.05 Ability to operate and monitor the following as they apply to an ATWS. BIT outlet valve switches. KIA MATCH ANALYSIS Question gives a plausible scenario with an ATWT in progress. Neither Boric Acid Transfer Pump is available. Candidate must choose a correct emergency boration flow path that would achieve Emergency Boration Flow. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Without BA Transfer Pumps available there would be no flow through HV-81 04. Plausible the candidate may not realize BA Transfer Pump impact on flow path. Flow rates given would satisfy the flow path if BA Transfer Pumps available. B. Incorrect. LV-112D and LV-112E would satisfy the flow path requirements but the minimum flow requirement via this path would be 100 gpm. Plausible candidate could recognize a correct flow path but confuse the flow rate requirements. C. Incorrect. FV-011 OA and FV-011 OB would not have flow through this path without the Boric Acid Transfer Pumps available. Plausible candidate may not realize BA Transfer Pump impact on the flow path and confuse the flow rate requirements. D. Correct. Opening LV-112D and LV-112E would establish boration flow from RWST and flow requirements would be satisfied with 100 gpm to BIT. 100 gpm used to sound more like choice B to make question symetrical with choices and NOT be a NOT question. REFERENCES 19211-C, Nuclear Power Generation ATWT page 4 13009-1/2, CVCS Makeup Control System section 4.9 for Emergency Boration pages 38 through 41. LO-PP-09300-06-001, 003, and 004 from Vogtle LO Active Exam Bank VEGP learning objectives: LO-PP-09300-06, Describe all emergency flow path a. borated water source and discharge flow path b. minimum flow requirements Page: 20f3 10/26/2009 Page: 30f3 10/26/2009

29. 032AK3.02 OOlINEW/RO/C/A 3.7/4.1/N/N/3/CVRIY Unit 1 is at 100%, and the following conditions occurred:

Intermediate Range Channel N-35 lost compensating voltage.
I&C is called to investigate.
Prior to any action by I&C, a reactor trip occurs. Which one of the following describes the Source Range NI detectors response after the trip, and the required actions lAW ESP-0.1, Reactor Trip Response?

Source Range Instruments will (1) ; and they must be manually (2) A. (1) automatically energize prematurely (2) de-energized until approximately 5 minutes post-trip -t-e-prevent-damage,-tQ...the,-_ B. (1) automatically energize prematurely (2) de-energized until approximately 15 minutes --aeteeter C. (1) NOT automatically energize when required (2) energized approximately 5 minutes post-trip-topreventa loss (1) NOT automatically energize when required (2) energized approximately 15 minutes post-trip ,te-pfevent'a-Iossofreactorpower' ,"i n GI Page: 75 of 200 12/1412009 A -Incorrect. Plausible, since examinee may believe loss of compensating voltage will make IR power read lower than actual and energize the Source range Nls above the power level that they are normally operated at. UOP-1.2 (step 5.18) & UOP-1.3 direct the Source Range Nls deenergized above P-6, IR>1 OE-1 0 amps, and applicant may believe there is similar guidance in ESP-0.1 for a premature energizing of the Source Range Nls. The second part is incorrect, since the decay into the Source range is at -1/3 dpm for about 6 decades, and thus takes about 15 minutes. Plausible, since confusion may exist between the decay into the source range from the power range and the limit on power ascension rate of 1 DPM from procedures to travel the required five decades. B -Incorrect. The first part is incorrect (see A). The second part is correct (see D). C -Incorrect. The first part is correct (see D). The second part is incorrect (see A). D -Correct. Loss of compensation is under compensated, which means that IR power will read higher than actual. SR automatic energization requires 2 of 2 IR detectors < P-6. Power drops immediately after a trip approximately one decade from 100% to approximately 7% (even though Nls indicate 0%). Then, it decays at approximately 1/3 dpm for 5 decades to 10E-10 amps in the IR. Assuming -1/3 DPM SUR for about 5 decades, -15 minutes post trip is when the Source Range is required, and automatically energized if both IR NI detectors are working properly. Per ESP-0.1, Step 12; since P-6 interlock should have reinstated the SR NI high voltage power, "verify source range detectors energized" is directed. "Verify" means take action to accomplish it if it didn't already happen. At this time, the other Intermediate range that is reading correctly will indicate that the power level is below P-6, and the source range Nls must be energized manually. ESP-O.1, Reactor Trip Response, Revision 29 12 Monitor nuclear instrumentation. 12.1 [CAl WHEN intermediate range indication less than 10-10 amps OR BYP & PERMISSIVE P-6 light off, THEN verify source range Page: 76 of 200 12.1 IF no source range detector energized, THEN within one hour verify adequate shutdown margin using FNP-1-STP-29.1, detectors -ENERGIZED SHUTDOWN MARGIN CALCULATION (TAVG 547??F), or FNP-1-STP-29.2, SHUTDOWN MARGIN CALCULATION (TAVG <547F OR BEFORE THE INITIAL CRITICALITY FOLLOWING REFUELING). 12/14/2009 Previous NRC exam history if any: 032AK3.02 032 Loss of Source Range Nuclear Instrumentation AK3. Knowledge of the reasons for the following responses as they apply to the Loss of Source Range Nuclear Instrumentation: (CFR 41.5,41.10 / 45.6 1 45.13) AK3.02 Guidance contained in EOP for loss of source-range nuclear instrumentation .. 3.7* 4.1 Match justification: This has a IR channel malfunction that causes the SR instruments to be de-energized at a time they should be energized and the procedural guidance and time when the SR instruments will be energized by the operator. Objective:

6. EVALUATE plant conditions and DETERMINE if any system components need to be operated while performing ESP-O.l, Reactor Trip Response. (OPS-52531B06)

Page: 77 of200 12/14/2009 04/03109 13 :42:27 FNP-1-UOP-2.1 5.21 WHEN the Source Range Permissive P-6 light goes off (2/2 intermediate ranges less than 1 x 10-10 amps), THEN check the following: 5.21.1 Check SOURCE RANGE TRAIN A TRIP BLOCED/HIGH VOLT OFF status light is OFF. 5.21.1 Check SOURCE RANGE TRAIN B TRIP BLOCED/HIGH VOLT OFF status light is OFF. ,(' 521 3 Check SR LOSS OF DETECTOR VOLAGE annunciator FA3 is clear. L range channel hi h voltage is NOT automaticall eY". r r l .... energIzed ue to an under compensate r malfunctioning if lJI\1-JP It/:J:::-. ¥-r intermediate range channel, THEN manually RESET the affected V-,yLAl source range high voltage when the operable intermediate range channel indicates less than 5 x 10-11 amps. __ \1--' r: . # ( 5.21.4.1 IF required, THEN take SOURCE RANGE BLOCK-RESET A TRAIN to RESET. __ 5.21.4.2 IF required, THEN take SOURCE RANGE W ( BLOCK-RESET B TRAIN to RESET. .22 5.23 _1-_1-WHEN the Source Range Nuclear Instruments are energized, THEN perform the following: 5.22.1 Verify the scaler timer aligned and operating properly. 5.22.2 Verify audio count rate amplifier aligned and operating properly. Within one hour after P-6 is reached perform the following: 5.23.1 Perform SR 3.3. 1. 1 (channel check) for the Source Range Nuclear Instruments. 5.23.2 Document the channel check in FNP-1-STP-1.0, OPERATIONS DAIL Y AND SHIFT SURVEILLANCE REQUIREMENTS. 5.23.3 IF the channel check cannot be performed, THEN verify adequate shutdown margin using FNP-1-STP-29.1, SHUTDOWN MARGIN CALCULATION (TAVG 547°F), or FNP-1-STP-29.2, (TAVG < 547°F OR BEFORE THE INITIAL CRITICALITY FOLLOWING REFUELING) Version 63.0 05112109 12:45:53 FNP-I-UOP-1.2 NOTE: The rod position corresponding to ECC 0.5% SK/K is determined in FNP-1-STP-29.6, Appendix 1, CALCULATION OF ESTIMATED CRITICAL CONDITION. 5.16 IF criticality has NOT been achieved with the rods withdrawn to 0.5% _1_ (500 pcm) past the estimated critical position, THEN performJhe following: _1-5.17 5.16.1 Insert all control banks to the bottom of the core. 5.16.2 Direct Chemistry to sample the RCS for boron concentration. 5.16.3 Re-calculate the ECC. 5.16.4 Determine and correct any discrepancy in the ECC. 5.16.5 IF no error can be found in the ECC or the RCS boron concentration, THEN contact the Reactor Engineering for assistance with the ECC. 5.16.6 IF required, THEN establish the correct critical boron concentration. 5.16.7 WHEN all errors have been corrected, THEN using the Inverse Count Rate Ratio Plot procedure per Appendix 1, withdraw the control rods in MANUAL to establish reactor criticality. tJcff!J, J f'-c Establish a startup rate of approximately 3/4 decade per minute. -. . .Y -,-,-,,=c...:.the Source Range Permissive P-6 light is on (112 intermediate ranges greater than 10-10 amps), THEN perform the following: ) 5.18.1 5.18.2 5.18.3 5.18.4 18.5 Block the Source Range High Flux Reactor Trip. [] Source Range BLOCK-RESET A TRN taken to BLOCK [] Source Range BLOCK-RESET B TRN taken to BLOCK On the Bypass and Permissive Panel, verify that the following windows are illuminated. [] SOURCE RANGE TRAIN A BLOCKED HI VOLTS OFF [] SOURCE RANGE TRAIN B BLOCKED HI VOLTS OFF Verify SR LOSS OF DET VOLTAGE annunciator F A3 is illuminated. Verify the Source Range NI drawers indicate zero voltage. Ensure the Scaler-Timer is shutdown per FNP-I-S0P-39.0, NUCLEAR INSTRUMENT A nON SYSTEM. Version 92.0 REACTOR SHUTDOWN AND RCS COOLDOWN A normal plant shutdown and cooldown are performed periodically for refueling or maintenance. Power reduction is performed by decreasing the external load on the turbine generator in conjunction with a boration of the RCS. This maintains control rod position and satisfies axial flux difference requirements and rod insertion limits. As power is decreased below 15%, the rods are put in manual control and the reactor operator manipulates the rods as necessary to control RCS temperature. When the turbine generator load has been decreased to approximately 50 MW the turbine is tripped and the control rods are positioned to maintain approximately 2% reactor power. RESPONSE TO A REACTOR TRIP The actions taken by reactor operators following a reactor trip are dictated by approved station procedures. These procedures ensure that the reactor is shut down, the turbine is tripped, normal and/or emergency power sources are available, and the plant response is as expected. If needed, compensatory actions are taken in accordance with the procedures. Figure 8-24 shows the behavior of a reactor power drop following a reactor trip. 100% A After power has been stabilized at ffi approximately 2%, the reactor operator records the information required for Reference YI).. d fM Reactivity Data (RRD): power level, rod B ,/ Y position, and actual boron concentration. This d.# I margins and for subsequent ECPs. Once ,-C data will be used for calculation of shutdownc;.- ___ c:w=-__ RRD data has been recorded, the reactor is .y>&"we-r ...v ____ . 0 shutdown by fully inserting all control banks. 0% L...-___________ --.......;;. __ Before starting the reactor cooldown, the RCS is borated to achieve the xenon-free shutdown margin required by technical specifications for RCS temperature below 200°F (typically -1,000 i1k1k). Once this boration is completed and the RCS boron concentration has been verified by chemical analysis, the cooldown is performed. When cold shutdown conditions are reached, the shutdown margin is re-verified. If adequate, the shutdown banks are fully inserted and the reactor trip breakers are opened. PWR / REACTOR THEORY / CHAPTER 8 / REACTOR OPERATIONAL PHYSICS ur w IUt::.lOUIll ..., 450[70 TIME AFTER TRIP Figure 8-24 Reactor Power Drop Following a r _ /) Reactor Trip Th1lfission rate decreases to below the power range immediately upon insertion of the control and shutdown rods. This rapid reactivity insertion is denoted by the neutron flux trace (power drop) from A to B in the figure. This is referred to as prompt drop following the reactor trip. During the period from B to C, the neutron population is dominated by the appearance of delayed neutrons from shorter-and intermediate-lived delayed neutron precursors. © 2007 GENERAL PHYSICS CORPORATION REV 4 VY '15J:' ,vllU v IUt::.lOUIll These precursors, which were formed when the reactor was at 100% power, decay within a few minutes. Once the shorter-lived precursors have effectively all decayed, neutron population is controlled by the appearance of delayed neutrons from the longest-lived precursors. From C to 0, power falls at a constant -80 second period based on the mean life of the longest lived delayed neutron precursor, bromine-87 (half-life of about 56 seconds). The -80 second period is eguivalent to about a -1/3 decade per minute (OPM) startup rate-(SUR). This continues until neutron popi:i1ai1Oil 18 low for the effect of source neutrons tobe seen and a subcritical eguilibrium is reached. --Core thermal power remains high for several seconds after the trip (as shown by points B to C). There is a time lag of a few seconds for the heat generated in the fuel to be conducted into the coolant, and the decay heat immediately following the prompt drop is approxImately 7% Of rated thermal power (R TP), assuming a trip 'trom equilibrium full power operation. (Thlli """occurs at about point C.)

RCS temperature is reduced by the steam dump system and stabilizes at no-load Tave.
Ten seconds after the trip, decay heat is still approximately 5% RTP, and it decreases to about 1% RTP in a little less than three hours (between points C to 0). PWR / REACTOR THEORY / CHAPTER 8 / REACTOR OPERATIONAL PHYSICS 460[70 A reactor that has been operating at state 100% power trips, dropping rods worth 1 0 (l0,000 pcm) into the core. This causes an immediate prompt drop in reactor power to approximately

___ %, followed by a slower decrease. I;? df?-M r--d eC -c.--:::;> Example 8-25 © 2007 GENERAL PHYSICS CORPORATION REV 4 05112/09 12:45:53 s o U R C E R A N G E Figure 1 I 10-3 N T E R 10-4 M E o I A 10-5 T E R A 10-6 N G E 10 6--.---10 5--1---10 4--1---10 3--1---10 2--1---10 1--1---10-7 10-8 10-9 10-1 0 10-1 1 --------Page 1 of 1 ------FNP-1-UOP-1.2 120% P 0 100% w---"-E I d.e-c "-c*e R t' lo\ f-t:v.A R J-\t7P 50% A L N G E 0% l-t, 1 -"2 -L '7 1 i --1 Version 92.0 05112/09 12:45:53 FNP-I-UOP-l.2 _1-_1-5.11 WHEN the reactor is critical, THEN perform one of the following: 5.11.1 5.11.2 Verify the low low Tavg alarm reset (RX COOLANT LOOPS lA, IB or lC TAVG LO-LO annunciator HF4) AND all RCS Loop Tavg greater than or equal to 547°F. Verify each reactor coolant loop Tavg greater than or equal to 541°F at least once ever 30 minutes per FNP-I-STP-35.l, UNIT STARTUP TECHNICAL SPECIFICATION VERIFICATION. (Technical Specification 3.4.2) CAUTIONS:

During all rod withdrawals, monitor nuclear instrumentation.

5.12 5.13 _1-Criticality shall be anticipated any time. the control rods are being withdrawn .. Additionally, during any approach to criticality monitor all pertinent instrumentation to allow errors in the ECC or problems with other instrumentation to be dete.cted early. Consider the use of audio count rate speakers as an aid to determine increasing flux rate. (SOER 88-02)

Do not exceed a sustained startup rate of one decade per mmute. Using the Inverse Count Rate Ratio procedure per Appendix 1 withdraw the control rod banks in MANUAL to establish reactor criticality.

{CMT-0008411} Verify proper overlap ofl.R. (See Fig 1) Note S.R. count when I.R. starts to come on scale. (CR 1-2000-148) 5.13.1 5.13.2 5.13.3 Record Source Range Counts. ____ (N31) ___ (N32) Record SR counts in Reactor Operators Log. Record SR counts in Surveillance Test Data Book. Version 92.0 FNP Units 1 & 2 REACTOR PROTECTION SYSTEM A-181007 2.8.1 Reactor Permissives

1. Power Escalation

2. The overpower protection provided by the excore nuclear instrumentation shall consist of three discrete, but overlapping levels. Continuation of startup operation of power increase shall require a permissive signal from the higher range instrumentation channels before the lower range level trips can be manually blocked by the operator.

Source range level trip block and high voltage cutoff shall always be active when above the permissive P-1 0 level. The intermediate range reactor trip and power-range (low setpoint) reactor trip shall only be blocked after satisfactory operation and permissive information are obtained from two of four power range channels which indicates P-1 O. Individual blocking switches shall be provided so that the low setpoint power range trip and intermediate range trip can be independently blocked. Moreover P-I0 allows the operator to manually block the intermediate range C-l rod stop. These trips are aut9matically reactivated when any three of the four power range channels are below the permissive (P-I0) level, thus ensuring automatic activation to more restrictive trip protection. See Table T-3 for a comprehensive list of Reactor Protection System permissives. (References 6.4.007, 6.4.011, 6.7.012) Blocks of Reactor Trips at Low Power Permissive P-7 shall prevent unnecessary at power reactor trips during low power by auto blocking the following reactor trips: Low reactor coolant flow in any two loops RCP breaker trip Undervoltage condition on RCP electrical buses Underfrequency condition on RCP electrical buses 2-32 Rev. 10 I

30. 035K3.02 005lNEW/ROICIA 4.0/4.3/N/N/4/HOWARDSIVER 5 EDITORIAL Unit 1 has experienced a Loss of Offsite Power and a Tube Rupture on the 1A SG, and the following conditions exist:

RCS cooldown at the maximum obtainable rate is in progress lAW EEP-3, Steam Generator Tube Rupture.
INTEGRITY Critical Safety Function Status Tree has turned ORANGE due to the 1A RCS LOOP cold leg temperature dropping rapidly. Which one of the following describes the reason the 1 A RCS LOOP cold leg temperature has dropped rapidly? 1A RCS Loop flow has _________

_ A. increased, moving the cold 1A SG U-tube water past the T COLD instruments. B. restarted, causing a sudden rise then rapid drop in temperature as the stagnant water from the hot leg is flushed through the loop. C. reversed, causing the cold water from 1 Band 1 C loops to pass over the T COLD instruments in 1A loop. stopped, allowing the cold Safety Injection water to pass over the T COLD instruments. Page: 78 of 200 12/1412009 A -Incorrect. During E-3 max rate cooldown under natural circ conditions, the ruptured SG loop flow will stagnate and may reverse, but NOT increase since the ruptured SG is not steamed, the differential temperature causing the Thermal Driving Head will be lost in that loop. Plausible: performing a cooldown increases the TDH for the intact loops and the cooldown for the intact loops would result in colder SG U-tube water to pass the T COLD instruments. B -Incorrect. SEE A. Loop flow is expected to stall not restart. Plausible: Initiating the cooldown, would restart or improve the intact SG loop flows. Also, after the termination of the cooldown/depressurization there is expected to be a minor restoration of flow in A RCS loop during the recovery actions and subsequent stabilization procedures. C -Incorrect. 1A RCS LOOP flow will stop, however T cold of the active loops will not be sufficiently low to cause integrity to be challenged in the inactive loop, otherwise the Integrity status tree would be VALID and thier temperatures would ALSO result in an ORANGE INTEGRITY condition. Plausible: a flow reversal is discussed in the occurrance of this condition and the 1 B & 1 C loops temperatures are lower than the 1A RCS loop. o -Correct. The basis for step 6.4 CAUTION-1 warns the operator to not enter FRP-P.1 if caused from the LOOP with the Ruptured SG. This is because SI flow reversal will likely occur in the ruptured Loop and "result in the indicated cold leg temperature (due to the location of the cold leg RTD)" to decrease. The flow stagnation in the 1A RCS loop, combined with the SI Flow into the loop, and a leak in the 1 A SG would cause an accumulation of the Cold SI (RWST) water to accumulate between the SI thermal sleeve and the SG, causing a >100°F/hr cooldown (which is in all loops due to operator action) AND <250°F (285°F unit 2)--an ORANGE path condition on FRP-P. EEB-3.0, ver 1, pg 31: Basis for step 6.4 CAUTION-1 "If the RCS is being cooled down on natural circulation during a steam generator tube rupture event, reverse flow through the ruptured loop during the cooldown or when the pressurizer PORV is opened to depressurize the ReS is possible and could cause the SI flow path in the ruptured loop to change. This change in the SI flow path could result in an indicated cold leg temperature (due to the location fo the cold leg RTD) that decreases to the point that the symptoms for FR-P.1 would occur. This false indication would only be seen in the ruptured loop since it is essentially stagnant while th either loops are circulated by natural circulation. When the PORV is closed, the flow paths are expected to change and the indicated cold leg temperature should increase resulting in the symptoms disappearing. When SI is terminated, the indicated cold leg temperature would increase if it did not do so earlier resulting in the symptoms for FR-P.1 no longer being present. This is an expected condition and the operator should only monitor the F-O.4, Integrity Status Tree for information purposes." Page: 79 of 200 12/1412009 Previous NRC exam history if any: None 035K3.02 035 Steam Generator System K3 Knowledge of the effect that a loss or malfunction ofthe S/GS will have on the following: (CFR: 41.7 145.6) K3.02 ECCS .......................................................... 4.04.3 Match justification: the effect of ECCS flow into the loop due to implementing E-3 is a stagnation/flow reversal in A loop, which is indicated by a rapid drop in indicated loop cold leg temperature. This rapid drop in temperature results in FRP-P.1 being ORANGE, and understanding the mitigation strategy IMPACT on ECCS flowpath ensures that the cooldown would not be erroneously terminating. Terminating the cooldown due to the indications presented here, would complicate stabilization of the plant. Objective: OPS-52530D03; State and Explain the basis for all Cautions, Notes and Actions associated with EEP-3 [ ... ]. Page: 80 of 200 12/14/2009 () j'\j .\i!! FNP-1-EEP-3 STEAM GENERATOR TUBE RUPTURE Revision 24 Step Action/Expected Response Response NOT Obtained n ************************************************************************************** CAUTION: With all RCPs secured RCS cooldown may cause a false FNP-1-CSF-0.4 Integrity Status Tree indication for the ruptured loop. Disregard ruptured loop cold leg temperature until completion of step 30. ************************************************************************************** NOTE:

The steam dumps will be interlocked closed when RCS TAVG reaches P-12 (543°F). This interlock may be bypassed for A and E steam dumps with the STM DUMP INTERLOCK switches .
Excessive opening of steam dumps can cause a high steam flow LO-LO TAVG main steam line isolation signal. 6.4 IF condenser available.

6.4 Dump steam to atmosphere. THEN dump steam to condenser from intact SGs at maximum attainable rate. BYP & PERMISSIVE COND AVAIL [] C-9 light lit STM DUMP [] MODE SEL A-B TRN in STM PRESS STM DUMP INTERLOCK [] A TRN in ON [] B TRN in ON STM HDR PRESS [] PK 464 adjusted 6.4.1 Direct counting room to perform FNP-0-CCP-645. MAIN STEAM ABNORMAL ENVIRONMENTAL RELEASE. 6.4.2 IF normal air available. THEN control atmospheric relief valves to dump steam from intact SGs at maximum attainable rate. IF NOT. dump steam using FNP-1-S0P-62.0. EMERGENCY AIR SYSTEM. 1A(1B.1C) MS ATMOS REL VLV [] PC 3371A adjusted [] PC 3371B adjusted [] PC 3371C adjusted Step 6 continued on next page. Page Completed Page 16 of 57 06127/0716:10:46 STEAM GENERATOR TUBE RUPTURE Plant Specific Background Information Section: Procedure FNP-0-EEB-3.0 Unit 1 ERP Step: 6.4 CAUTION-1 Unit 2 ERP Step: 6.4 CAUTION-1 ERG Step No: 6 CAUTION-I ERP StepText: With all RCPs secured RCS cooldown may cause a false FNP-2-CSF-0.4 Integrity Status Tree indication for the ruptured loop. Disregard ruptured loop cold leg temperature until completion of step 30. ERG Step Text: If RCPs are not running, the following steps may cause a false F-O.4, Integrity Status Tree indication for the ruptured loop. Disregard the ruptured loop T-cold indication until after performing Step 29. Purpose: To alert the operator that during a natural circulation cooldown a false symptom of a red or orange path condition in F-O.4, Integrity Status Tree is possible due to the redirection of SI flow in the ruptured loop. Basis: If the RCS is being cooled down on natural circulation during a steam generator tube rupture event, reverse flow through the ruptured loop during the cooldown or when the pressurizer PORV is opened to depressurize the RCS is possible and could cause the SI flow path in the ruptured loop to change. This change in the SI flow path could result in an indicated cold leg temperature (due to the location of the cold leg RTD) that decreases to the point that the symptoms for FR-P.1 would occur. This false indication would only be seen in the ruptured loop since it is essentially stagnant while the other loops are circulating by natural circulation. When the PORV is closed, the flow paths are expected to change and the indicated cold leg temperature should increase resulting in the symptoms disappearing. When SI is terminated, the indicated cold leg temperature would increase if it did not do so earlier resulting in the symptoms for FR-P.1 no longer being present. This is an expected condition and the operator should only monitor the F-O.4, Integrity Status Tree for information purposes. After the cooldown and depressurization is completed and SI is terminated, the operator should monitor the F-O.4, Integrity Status Tree to determine if a red or orange path still exists and FR-P.1 should be implemented. His decision should be based on the symptoms existing after SI is terminated. If a multiple or subsequent accident occurs, the operator could transfer out of E-3 prior to terminating SI. For that case he should monitor the F-O.4, Integrity Status Tree when he makes the transition out of E-3 to determine if at that time a red or orange path exists and FR-P.1 should be implemented. STEP DESCRIPTION TABLE FOR E-3Step 6 -CAUTION Knowledge: If a multiple or subsequent accident occurs, the operator could transfer out of E-3 prior to terminating SI. For that case he should monitor the F-O.4, Integrity Status Tree when he makes the transition out of E-3 to determine if at that time a red or orange path exists and FR-P.1 should be implemented.

References:

DW 028 31 of 119 Version: 1.0

31. 037AA2.0S OOlINEW/RO/C/A 2.S/3.3/N/N/2/CVRlY Unit 1 is at 12% power, and the following conditions exist:

R-15A, SJAE EXH, has failed.
1 A SG has developed a 10 gpm tube leak.
One of the 1A SG safeties is leaking by. Which one of the following radiation monitors will provide the EARLIEST indication of the 1A SG Tube leak? A'I R-19, SGBD SAMPLE, alarm. B. R-23B, SGBD TO DILUTION, alarm. C. R-70A, 1A SG TUBE LEAK DET, alarm. D. R-60A, 1A STEAM GENERATOR, alarm. A -Correct. R-19 is continuously monitoring the SGBD system sample stream and will be the first indication of an alarm considering only the 4 choices given. B -Incorrect.

R-23B will only alarm after the SGBD surge tank starts filling with the contaminated SGBD water. The tank is maintained 50% full, and there will be a diluting effect at first. Downstream of this tank is R-23B in a flow stream going to the environment. R-19 samples undiluted SGBD water continuously, and would alarm sooner than R-23B. Plausible, since R-23A samples blowdown water at the inlet of the Surge Tank and is undiluted SGBD water. Due to the higher flowrate of SGBD (about 130 gpm) than the sample stream, it alarms sooner than the R-19 alarm for a particular SGTL event. Confusion may exist as to the difference between the choice for R-23B and R-23A which would alarm prior to R-19 (as seen on the simulator during SGTL events). C -Incorrect. R-70A alarm setpoints are not valid at this power level. The R-70s shift automatically from the gpd Mode to the ME mode below 20% power, and the alarm functions are set for gpd. Plausible, since it is on the Steam line at the outlet of the SG, and alarms first before any other Radiation monitor in the event of a SGTL above 20% reactor power. D -Incorrect. R-60 is a high range monitor that does not upscale in the event of SGTL with no fuel failure, even with an open safety or SG Atmospheric relief. Plausible, since if it was a lower scale, it would alarm prior to R-19 since it is monitoring the steam coming from the 1A SG with the tube leak. Also, it would alarm if the dose from the SG was high enough (such as due to a SGTR and fuel element failure). The SGTL combined with the safety leakby allows it to monitor the actual1A SG contaminated steam as it escapes through the safety. FNP-I-AOP-2.0, Steam Generator Tube Leakage, Version 33.0 B. Symptoms or Entry Conditions I. Enter this procedure when RCS tube leakage is indicated by high secondary activity on any of Page: 81 of200 12/1412009 the following radiation monitors or by sample results. a. R-15 SJAE EXH [listed here to show the nomenclature in the procedure]

b. R-15B or R-15C TURB BLDG VNTL c. R-19 SGBD SAMPLE [listed here to show the nomenclature in the procedure]d.

R-23A SGBD HX OUTLET e. R-23B SGBD TO DILUTION [listed here to show the nomenclature in the procedure]

f. R-70A, R-70B or R-70C IA(IB,IC)

SG TUBE LEAK DET g. SG sample results indicate primary to secondary leakage for any SG greater than or equal to the normal alarm setpoint for annunciator FGI, SG TUBE LEAK ABOVE SETPT. FNP-l-ARP-l.6, FHl AUTOMATIC ACTIONS (cont) 2. ARDA will automatically start for the following conditions: 2.1 ARDA will automatically start when any of the following monitors go into alarm for two consecutive system polls one minute apart on either unit and use the latest 15 minute average monitor value to perform the calculations: Plant Vent Stack Monitors R29 (SPING) Noble Gas 4.44e-4 clml Iodine 1.20e-6 clml Particulate 4.00e-5 clml Steam Jet air Ejector R15C 27 mrlhr TDAFW Exhaust R60D 38 mrlhr Steam Generator A R60A 38 mrlhr [listed here to show the nomenclature in the procedure] Steam Generator B R60B 38 mrlhr Steam Generator C R60C 38 mrlhr Page: 82 of 200 12/1412009 Previous NRC exam history if any: 037AA2.08 037 Steam Generator Tube Leak AA2. Ability to determine and interpret the following as they apply to the Steam Generator Tube Leak: (CFR: 43.5 145.13) AA2.08 Failure of Condensate air ejector exhaust monitor ..................... 2.83.3 Match justification: A scenario is given with a SGTL and a failed Condensate air ejector exhaust monitor (R-15A), which is normally the first indication of a SGTL. To answer this question correctly, determining how the failed R-15 applies to the SGTL is required. I. E., since it is no longer the first indicaiton of a SGTL, which is the first indication? Objective:

2. RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the components associated with the Radiation Monitoring System to include those items in Table 4-Remote and Local Indications and Controls (OPS-40305A02).

5. DEFINE AND EVALUATE the operational implications of normal/abnormal plant or equipment conditions associated with the safe operation of the Radiation Monitoring System components and equipment, to include the following (OPS-40305A07):

Normal control methods
Abnormal and Emergency Control Methods
Automatic actuation
Protective isolations
Protective interlocks
Actions needed to mitigate the consequence of the abnormality Page: 83 of 200 12/1412009 03112/03 11 :24:25 1.0 Purpose FARLEY NUCLEAR PLANT UNIT 1 SYSTEM OPERATING PROCEDURE SOP-69.0 N-I6 PRIMARY TO SECONDARY LEAK DETECTION SYSTEM FNP-I-SOP-69.0 To provide guidance for operation of the Primary to Secondary Leak Detection System. 2.0 Initial Conditions 2.l 120V Regulated Instrumentation PaneliB is energized per FNP-I-SOP-36.4, 120V A.C. DISTRIBUTION SYSTEMS. 3.0 Precautions and Limitations 3.1 The system receives a reactor power input from power range channel N-43. IF .

fails OR is in Test OR is less than 20% power, THEN the system cannot C 1 \\&7 f accurately estimate a leak rate in the A V mode, and the indicators will display oJ /7 "PN <20%". If desired, the Counting Room can configure the N-16 system in the ')0/ ME counts per second (CIS) mode using FNP-0-CCP-31, LEAK RATE Q DETERMINATION. While not able to provide a leak rate determination, this /\Jt' mode can be used to indicate if leakage is increasing based on the indication trending up. The A V mode is the preferred mode of operation above 20% reactor power. The ME mode should only be utilized below 20% reactor power. 3.2 The N -16 Leak Detection System cannot determine the location of a leak within a specific Steam Generator. The system can however provide a more accurate leak rate determination if the location of the leak is known to be in one of the following locations: Cold Leg -CB, Hot Leg -HB or U-Bend region -BE WHEN a leak location is selected (CB, HB or BE), THEN the processor displays a leak rate that assumes the leak is at the location you have selected. The A V mode is essentially the average of the three leak rates at the specific locations. 3.3 The N-16 system is limited to an upward range of 1,000 gallons per day. Version 5.0 FNP Units 1 & 2 RADIA TION MONITORING SYSTEM A-181015 increasing radiation to initiate the RMS High Radiation annunciator on the main control board. In addition, each function shall actuate an indicating light on the ratemeter front panel defining the alarm condition (References 6.4.034,6.4.214, and 6.4.249). The ratemeter operation selector switch shall actuate the RMS CH Test annunciator on the main control board when placed in any position other than OPERATE (References 6.4.034 and 6.4.249). 3.2.5.3.4 Normally, no contaminated leakage is expected into the Service Water system. Accordingly, the monitor setpoint should be set approximately one half decade above the detector's normal response (Reference 6.7.062 and 6.7.080). 3.2.5.4 Interface Requirements The instrument power supply for the RMS system panel NIHIINGRM 2502A, B, and C is 120 VAC distribution panel IB, breaker number 2 (Reference 6.4.219). The control power supply for the RMS system panel NIHIINGRM 2502A, B, and Cis 2081120 VAC control power panel IN, breaker number 6 (Reference 6.4.107). The instrument power supply for the RMS system panel N2HIINGRM 2502A, B, and Cis 120 VAC distribution panel2B, breaker number 2 (Reference 6.4.345). The control power supply for the RMS system panel N2HIINGRM 2502A, B, and C is 2081120 VAC control power panel 2N, breaker number 6 (Reference 6.4.106). 3.2.5.5 Shielding Design In addition to the shielding provided by the monitor housing, additional shielding was added surrounding the high voltage electronics housing above the detector. The purpose of the added shielding is to reduce the influence of background radiation, that caused spiking of the monitors and isolation of the process system and to improve monitor sensitivity (References 6.7.033, 6.7.035, and 6.7.080). 3.2.6 Steam Generator Blowdown 23162.3431A-181015.RM Service SG Blowdown to Processing System SG Blowdown Discharge 3-34 TPNS Nos. NDllRE 0023A NDllRE 0023B Rev. 0 FNP Units 1 & 2 RADIATION MONITORING SYSTEM A-181015 23162.3431A-181015.RM 3.2.6.1 Basic Function 3.2.6.1.1 Radiation detector RE 0023A, located in the steam generator blowdown discharge line upstream of the steam generator blowdown surge tank, monitors for an increase in radioactivity in the secondary system. To minimize contamination of the processing system and potential inadvertent release of radioactive gases through the surge tank vent, steam generator blowdown is automatically terminated by closing valve NB21 FCV 1152 when the activity exceeds the setpoint. An increase in radioactivity in the secondary system would be indicative of a steam generator tube rupture accident (References 6.7.084 and 6.4.366). 3.2.6.1.2 Radiation monitor RE 0023B is located downstream of the steam generator blowdown discharge pumps prior to discharging to the dilution discharge on the service water system. This detector monitors the discharge stream to comply with GDCs 60 and 64. On an increase in radioactivity, the discharge is isolated by automatically closing NB21RCV023B (References 6.7.084 and 6.4.366). 3.2.6.2 Functional Requirements 3.2.6.2.1 An in-line liquid monitor shall be provided to directly monitor the process medium. The use of this type of monitor provides the fastest response time and easiest decontamination (References 6.4.366 and 6.7.080). 3.2.6.2.2 RE 0023B shall alarm and isolate the effluent discharge prior to exceeding the limits of ten times the concentrations stated in 10 CFR Part 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in unrestricted areas will result in exposures within (1) the Section ILA design objectives of Appendix I, 10 CFR 50, for a member of the public, and (2) the limits of 10 CFR 20.1301 for the population. The limiting concentration of dissolved and entrained noble gases is 1 x 10-4 !-LCi/ml (Reference 6.7.078, 6.7.081). Setpoints are based on ensuring the discharge limits presented in Section 2.1.2 of the ODCM are not exceeded. 3-35 Rev. 0 FNP Units 1 & 2 RADIA nON MONITORING SYSTEM A-181015 3.3.16.5 Interface Requirements 3.3.16.5.1 The 120 VAC power supply for RMS panel QSD11RE 0035A is 2081120 VAC control power panel 1R, breaker number 15 (Reference 6.4.1 08). The 208 V AC pump power supply for RMS panel QSD11RE 0035A is 2081120 VAC control power panel 1R, breaker number 12 (Reference 6.4.108). The 120 VAC power supply for RMS panel QSD11RE 0035B is 2081120 VAC control power panel 1 S, breaker number 15 (Reference 6.4.108). The 208 V AC pump power supply for RMS panel QSD lIRE 0035B is 2081120 VAC control power panel IS, breaker number 12 (Reference 6.4.108). 3.3.16.5.2 The instrument air system shall provide a dry, filtered air source for monitor purge air (Reference 6.7.080). 3.3.17 Main Steam Safety and Atmospheric Relief and TDAFW Pump Exhaust Noble Gas Monitors 23162.3431A-18101S.RM Service Steam Generator A Safety!Atmospheric Relief Valve Exhaust Steam Generator B Safety! Atmospheric Relief Valve Exhaust Steam Generator C Safety! Atmospheric Relief Valve Exhaust Turbine Driven AFW Pump Exhaust 3.3.17.1 Basic Function TPNS Nos. NDllRE 0060A ND11RE 0060B ND 11 RE 0060C ND 11 RE 0060D 3.3.17.1.1 The monitors provide post-accident effluent monitoring for the Steam Generator Safety Valve and Turbine Driven Auxiliary Feedwater Pump turbine exhaust points in compliance with RG 1.97 (References 6.4.051 and 6.4.350). 3.3.17.1.2 These monitors may be used for RG 1.21 effluent activity tracking (References 6.7.078 and 6.7.060). 3.3.17.2 Functional Requirements 3.3.17.2.1 The monitors shall provide continuous indication over a range of 10-1 to 10 3 microcuries per cubic centimeter 3-93 Rev. 0 FNP Units 1 & 2 RADIATION MONITORING SYSTEM A-181015 (llCilcc). The monitors shall provide an approximately linear response for gamma energies between 0.5 and 3 MeV (References 6.4.051 and 6.4.350, and 6.7.003). NUREG 0737, Clarification Item Il.F.1, Attachment 1, requires that noble gas radiation monitors be provided for effluent points which monitor from normal operating levels to a maximum of 10 5 llCi/cc (Xenon-133 calibration) for undiluted containment effluents and lO-1 to 10 3 llCi/cc for buildings with systems containing primary coolant such as the auxiliary building. The present plant configuration provides area monitors for these parameters used in a process monitoring application. The meters for these monitors read in R/hr and conversion charts have been provided to convert the meter readings to llCi/cc. The present monitors indicate a range of 10-2 to 10 6 mR/hr, which corresponds to 10-5 to 1.4. x 10 4 llCi/cc (References 6.3.001 and 6.7.005). 3.3.17.2.2 Each radiation monitor is located to view its respective steam generator safety valve plumes and atmospheric relief valve plume. Each monitor is located on the auxiliary building roof and oriented to minimize the effects of radiation shine from the containment following a design basis LOCA. The monitors are located as far from the safety valve and main steam atmospheric vent valve vent stacks as permitted by the containment wall while keeping all monitors the same distance from the centerline of their respective main steam atmospheric vent valve. The monitor for the "B" steam generator limits this distance to 17.17 feet. Therefore, the monitors are located on a 17.17-foot circle around the main steam atmospheric vent valve stack and in a configuration not facing the containment, but have the steam generator safety valve plumes and the main steam atmospheric vent valve plume for that particular steam generator in full view. The viewing angle in the horizontal plane was determined based on the location of the monitors and was determined to be 55°. A viewing angle of 55° allows monitoring of all plumes of a particular steam generator while excluding the field of view from most of the other steam generator plumes, and not facing containment. For the vertical plane, the monitor viewing angle must be small enough to prevent the monitor from being influenced from radiation shine from the containment and large 3-94 Rev. 1

32. 038EA2.07 OOllNEW/ROICIA 4.4/4.8ININ/3/CVRlY Unit 1 has manually Tripped and Safety Injected from 14% power lAW AOP-2, Steam Generator Tube Leakage. The following conditions exist:

DA-07, 1A 4160V BUS SUPP FROM 1A S/U XFMR, breaker tripped open.
AFW Flows were maintained matched to all 3 SGs until securing AFW Flow.
AFW Flow has been secured to all SGs.
SG Pressures:
SG NR levels: -1A 980 psig and stable -1A 61 % and stable -1 B 980 psig and stable -1 B 61 % and rising -1 C 980 psig and stable -1 C 50% and lowering Which one of the following correctly describes the event in progress based on the MCB indications?

A'I SGTR on 1 B SG ONLY. B. SGTR on 1A AND 1 B SGs ONLY. C. SGTR on 1 B SG AND a Steam Leak on 1 C SG ONLY. D. SGTR on 1A AND 1 B SG AND a Steam Leak on 1C SG. A -Correct. The level rising in 1 B SG with no AFW flow, with max chg and no letdown and all SG pressures stable indicates this answer to be correct. C SG Ivl is decreasing due to being the only SG steaming with no AFW flow. This is normal indication for this condition, and will require AFW flow to maintain C SG Ivl. B -Incorrect. The SGTR on 1 B SG is correct, but the SGTR on 1 A is incorrect. Plausible, since with the AFW flows matched to all SGs, and 1A SG level higher than the other intact SG by 10% NR Lvi, a SG tube leak would be indicated if not for the tripped RCP in that loop, and the Ivl being stable instead of rising with no AFW flow AND no steaming. If the RCP was not tripped, this would be a correct answer, since with no AFW flow and decay heat removal level staying constant would indicate a SGTR. C -Incorrect. The SGTR on 1 B SG is correct, but the Steam Leak on 1 C SG is incorrect due to the pressure being stable at 980 psig on all three SGs. Plausible, since 1 C Level is dropping with the 1 A SG level stable and the 1 C SG dropping and no AFW flow to any of the SGs. D -Incorrect. SGTR on 1A & 1 B incorrect (see B). Steam leak on 1 C incorrect (see C). Page: 84 of 200 12114/2009 Previous NRC exam history if any: 038EA2.07 038 Steam Generator Tube Rupture EA2 Ability to determine or interpret the following as they apply to a SGTR: (CFR 43.5/45.13) EA2.07 Plant conditions, from survey of control room indications .............. 4.4 4.8 Match justification: Control Room indications are given which could indicate a SGTR in two SGs and a Steam leak in one SG under slightly different conditions than given. With one tripped RCP one SG level is high due to not steaming and not due to a SGTR. One SG is high due to a SGTR. One SG level is low and dropping due to being the only intact SG producing steam, even though the other intact SG Ivl is stable (due to the tripped RCP). The applicant must correctly evaluate all these indications and diagnose the event to be a SGTR in one SG only. Objective:

3. LIST AND DESCRIBE the sequence of major actions, when and how continuous actions will be implemented, associated with (1) EEP-O, Reactor Trip or Safety Injection and (2) ESP-O.O, Rediagnosis. (OPS-52530A04)

Page: 85 of200 12/1412009

33. 039K4.05 OOl/FNP BANK/RO/MEM 3.7/3.7/N/N/2/CVRIY Which one of the following adequately describes the setpoint of the steam line flow for the High Main Steam Line Flow with Low-Low T avg MSIV isolation?

A. Increases linearly from 40% to 110% steam flow as power increases from 0% to 100%. B. Increases linearly from 20% to 110% steam flow as power increases from 0% to 100%. C. Constant 20% steam flow up to 10% power; then increases linearly to 110% flow as power increases from 10% to 100%. Constant 40% steam flow up to 20% power; then increases linearly to 110% flow as power increases from 20% to 100%. A -Incorrect. Plausible, since the numbers are the same as the values for the correct setpoint, but the Constant value of 40% steam flow limit from 1 %-20% is left out of this choice. B -Incorrect. Plausible, since the numbers are the same as the values for the correct setpoint, but the Constant value of 40% steam flow limit from 1 %-20% is left out of this choice. C -Incorrect. Plausible, since this choice correctly states that there is a constant value of Steam Flow setpoint up to a certain power, but the constant value and associated power level are incorrect. The setpoint at 100% power is correct. 0-Correct. OPS-52201K A higher than expected steam flow from the steam generators, along with a decreasing T avg, is another indication of a steam break that will shut the MSIVs. The high steam flow set point is varied with turbine power by Pimp. The set point is 40 percent steam flow from 0 percent to 20 percent turbine power. It then increases linearly from 20 percent turbine power to 100 percent turbine power where the set point is 110 percent steam flow. 2/3 steam lines reaching the set point and T avg below the P-12 set point will shut the MSIVs. It requires only one of the two, density-compensated steam flow detectors per steam line to reach the set point to actuate the MSIV closure with a one second time delay. This main steam line isolation is not able to be blocked or bypassed. Page: 86 of 200 12/1412009 Previous NRC exam history if any: 039K4.05 039 Main and Reheat Steam System K4 Knowledge ofMRSS design feature(s) andlor interlock(s) which provide for the following: (CFR: 41.7) K4.05 Automatic isolation of steam line ................................... 3.7 3.7 Match justification: Other MSIAS are on exam so this type of question was selected to avoid double jeopardy. Objective:

6. DEFINE AND EVALUATE the operational implications of normal/abnormal plant or equipment conditions associated with the safe operation of the Main and Reheat Steam System components and equipment, to include the following (OPS-40201A07):

Normal control methods
Abnormal and Emergency Control Methods
Automatic actuation including setpoint (example SI, Phase A, Phase B, MSLIAS, LOSP, SO level)
Protective isolations such as high flow, low pressure, low level including setpoint
Protective interlocks
Actions needed to mitigate the consequence of the abnormality Page: 87 of 200 12/14/2009

5. High-1 containment pressure (pB-95IB, PB-952B, PB-953B) 4.0 psig "'" 6. Time delay on SI manual reset 1 minute B. Steam line isolation 1a. High steam line flow in coincidence with 10-10 T..-a dip corresponding to (FB-474A, FB-475A, FB-484A, FB-485A, 110% offull steam flow FB-494A, FB-495A) at full load 40% of full steam flow between 0% and 20% load dip setpoint linear with turbine first stage pressure between load and full load lb. 1..0-10 Tava (P-12) (TB-412E, TB-422E, TB-432E) 543 0 F Ie. Filter Lag Time Constant (FY-474B, FY-475B, FY-484B, o Seconds* FY -485B, FY -494B, FY -495B) *May be set up to 1.5 seconds 2. Low steam line pressure (pB-474A, PB-485A, PB-496A) SIS psis Lead time constant (pY -474B, PY -485B. PY -496B) SO seoonds Lag tUne constant (pY -474B. PY -485B, PY -496B) S seccmds 3. High-2 containment pressure * (pB-95IB, PB-952B, PB-953B) 1.6.2 psig Rev.A7 11

34. 041K4.14 00l/NEW/ROIC/A 2.S/2.8/N/N/4/CVRN Unit 1 was at 26% power and 180 MWe, and the following conditions occurred:

The reactor tripped.
The "A" Reactor Trip Breaker failed to open. Which one of the following correctly states the arming signal for the Steam Dumps, and the RCS temperature maintained by the Steam Dumps? The Steam Dumps are armed due to the (1) and RCS temperature will be controlled at (2) (1 ) (2) A. P-4 signal 54rF B. P-4 signal 551°F CY' Loss of Load signal 54rF D. Loss of Load signal 551°F A -Incorrect.

The first part is incorrect but plausible, since the reactor did trip and if A RT bkr would have opened this would be correct. Most functions of the P-4 Permissive come from both trains, but this function comes only from A train. The second part is correct, and is the result of the B train P-4 signal which is present as normal with the B train RT bkr open. B -Incorrect. The first and second part are incorrect but plausible, since this choice would be correct for a B train RT bkr failing to open. A train P-4 would arm the Steam Dumps and the LOL controller would stay in the circuit (since the B train P-4 did not shift controllers to the Plant Trip mode) to control Tavg 4°F higher than no load Tavg (547+4=551). C -Correct. The A train P-4 did not arm the steam dumps, and the Loss of Load did (the loss of load was 20% instantaneously, and thus greater than the LOL arming setpoint of 15% with a 120 secono time constant). The B train P-4 shifted the controllers from the LOL 1.9 which maintains a constant no load Tavg of 54rF. D -Incorrect. The first part is correct (see C). The second part is incorrect, but plausible, since it would be correct for a loss of load or reactor trip with the B train RT bkr open instead of the A. Page: 88 of 200 12/1412009 Previous NRC exam history if any: None 041 K4.14 041 Steam Dump System and Turbine Bypass Control K4 Knowledge of SDS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) K4.14 Operation of loss-of-load bistable taps upon turbine load loss ........... 2.5

2.8 Match justification:

The loss-of-bistable arms the steam dump when the loss of load magnitude is greater than the variable setpoint in a given time. The Reactor trip overrides the arming of the loss of load due the A RT bkr P-4 signal for a normal reactor trip. This question requires knowledge the loss of load setpoint and the times that it does and does not arm the Steam dumps. It also requires knowledge of what controller (Plant Trip or Loss of Load) is in the circuit under different conditions than normal. Several design features and interlocks relating to and affecting the Loss of Load interlock must be understood to correctly answer this question. Objective:

5. DEFINE AND EVALUATE the operational implications of abnormal plant or equipment conditions associated with the operation of the Steam Dump System components and equipment to include the following (OPS-5220 1 G07):

Normal Control Methods (Steam dump valves)
Abnormal and Emergency Control Methods (Steam dump valves, Steam dump Page: 89 of 200 system solenoid-operated three-way valves)
Automatic actuation including setpoint (High-1 and High-2 trip bistables)

Protective isolations (Plant trip controller, Loss of load controller, C-7)

Protective Interlocks (Condenser available, C-9, Low-Low T A VG signal, P-12) Actions needed to mitigate the consequence of the abnormality 12/1412009 Date: 10r '0009 ,-i I +rY BY , ..... OTHERS r-REDUNDANT I '-------,\: ---------L ----I I CONDENSER AVAILA1LE SIGNALS STEAM DUMP I 1/ A , INTERLOCK SELECTOR I I CIRCULATION

\lATER S'w'lTCH (NOTE 3) I CONDENSER PUMP PRESSURE S'w'ITCH BREAKERS CLOSED , ) REDUNDANT IBY IOTHERs L -------------REDUNDANT ------BY I (l) BY InTHFRs, MEDIAN T AVG <SHEET 9) REFERENCE T AVG INTERNAL SETPDINT Time "\44:26 PM 1 STEAM DUMP CONTROL MODE SELECTOR S,,",JTCH --1------.., , , , , , , , , , , , , , , , , , , , r , , , r , , (l) ,OTHERS STEAM HEADER PRESSURE , I r , , , , r , , , , , , r , , r , , , r , , , (l) IOTHERS; STEAM LINE PRESSURE MODULATE MODULATE MODULATE ATMOSPHERIC ATMOSPHERIC ATMOSPHERIC f NOTES: 1. THE LOOP I THE LOOP 2 THE LOOP 3 I ! RELIEF RELIEF _VALVE .OTHERS STEAM DUMP IS BLOCKED BY BLOCKING AIR TO THE DUMP VALVES AND VENTING THE DIAPHRAGM. THE , , , , , r , , , r , , r , tJ REDUNDANT LOGIC OUTPUT OPERATES 2 SOLENOID VENT VALVES IN SERIES TO REDUNDANTLY INTERLOCK THE AIR LINE BET\.IEEN EACH VALVE DIAPHRAGM AND ITS ASSOCIATED POSITIONER. THE SOlENDlD VALVES ARE DE -ENERGIZED TO VENT CAUSING THE MAIN DUMP VALVE TO CLOSE IN FIVE SECONDS, 2. CIRCUITRY ON THIS SHEET IS NOT REDUNDANT EXCEPT \.IHERE INDICATED REDUNDANT.

3. SELECTOR S\.IITCH \.lITH THE FDLLO\.lING 3 POSITIONS, ON-STEAM DUMP IS PERMITTED BYPASS-T AVG. INTERLOCK IS BYPASSED FOR LO\.l T.AVG. SPRING RETURN TO ON POSITION.

orr -STEAM DUMP IS NOT PERMITTED AND RESET T.AVG, BYPASS. THE REDUNDANT INTERLOCK SELECTOR S\.IITCH CONSISTS OF T\.IO CONTROLS ON THE CONTROL BOARD, ONE FOR SIGNAL INPUTS COMING FROM TURBINE PRESSURE MUST COME FROM DIFFERENT PRESSURE TAPS TO MEET THE SINGLE FAILURE CRITERION. TO 5. THE CONDENSER AVAILABLE SIGNAL LOGIC IS TYPICAL, VALVES MODULATED OPEN DR ACTUAL IMPLEMENTATION MAY BE DIFFERENT. CLOSED <ZERO TO FULL OPEN) 6, ALL TEMPERATURE BIST ABLES ON THIS SHEET AND TURBINE N1N36V501 N1N36V501E IMPULSE CHAMBER PRESSURE BISTABLE tlPB-447A NIN36V501G ARE TNERGIZE TO ACTUATE.6 NIN36V501B, NIN36V501F

7. LIGHTS SHOULD BE PROVIDED IN THE CONTROL ROOM FOR NIN36V50lD, NIN36V501H EACH DUMP VALVE TO INDICATE 'WHEN THE VALVE IS FULL Y CLOSED OR FULLY OPEN. (§ U166240.DVG ... J GaR !Z-Hl

@ C) o ..q-00..1 Zc..o I ::J ; nn -1.1 TIM AlA.6.bJ>lA. Pr::l'W'E.R COM?A."-l'1' ;m NO.1 I Z IMll!' i ... -r; FARLEY U-166240 ; I sv .... xi!lA'5 a. 4.. If)-' 2° ",,0:: ",z ",,0 6° ",,2 z:::> OD zw "-If) Title: G:\U166240.cal FNP Units 1 & 2 REACTOR PROTECTION SYSTEM A-181007 range channels exceeds a current equivalent to 20 percent reactor power. The rod stop may be manually blocked when above the P-l 0 setpoint, but is automatically reinstated below P-I0 (3/4). This can be manually bypassed at NIS racks. (References 6.4.007, 6.4.011, 6.4.016) C-2 Interlock. The C-2 overpower rod stop blocks automatic and manual control rod withdrawal. The block action occurs when 114 power range channels exceeds 103 percent reactor power. Each power range channel may be manually bypassed at the NIS racks. All power range channels cannot be bypassed at the same time. Only two power range channels may be blocked at one time using two switches located on the NIS Miscellaneous Control and Indication Drawer. (References 6.4.007, 6.4.011,6.4.016) C-3 Interlock. The C-3 control interlock is generated by the OTDT circuitry. The setpoint is 3 percent below the variable OTDT reactor trip setpoint. C-3 generates a block of automatic and manual rod withdrawal, when 2/3 loop delta Ts exceed their setpoint. The function of the rod block is to eliminate the cause of the impending trip, thereby preventing it. Since relatively slow transients are typical of those requiring OTOT protection, there is sufficient time for a load reduction to correct the situation. (References 6.2.003, 6.4.007, 6.4.012, 6.4.016) C-4 Interlock. The C-4 control interlock is generated by the OPDT circuitry. The setpoint is 3 percent below the variable OPOT reactor trip setpoint. C-4 generates a block of automatic and manual rod withdrawal, when 2/3 loop delta Ts exceed their setpoint. The function of the rod block is to eliminate the cause of the impending trip, thereby preventing it. (References 6.2.003, 6.4.007, 6.4.012, 6.4.016) C-5 Interlock. The C-5 interlock ensures that automatic rod withdrawal system is prevented when less than 15 percent power. It also prevents automatic rod withdrawal when power falls below 15 percent. The setpoint is 15 percent power as detected by turbine first stage impulse pressure. (References 6.4.007,6.4.016,6.4.022) C-7 Interlock. This control interlock arms the steam dumps upon a load rejection (when in coincidence with C-9). The steam dump demand interlock (C-7) is actuated when turbine load is reduced by greater than 15 percent with a 120 second time constant. Rate differentiation of the first stage turbine impulse chamber pressure signal provides the equivalent turbine load signal. It must be manually reset. (Only PT-447 provides input to C-7.) (References 6.4.007, 6.4.017) C-9 Interlock. C-9 is the condenser-available interlock. This interlock allows the steam dump valves to be armed if the condenser is available. It also prevents an overpressure condition which could damage the 2-36 Rev. 10 I FNP Units 1 & 2 REACTOR PROTECTION SYSTEM Pressurizer low pressure Pressurizer high level A-I8I007 Permissive P-7 shall block the above listed reactor trips below 10 percent of full power. The low power signal is derived from the power range neutron flux channels (P-I 0) and the turbine impulse chamber pressure channels (P-13). The blocking feature occurs during the absence of P-7, meaning that P-7 is not active. This occurs when 3/4 power range neutron flux channels are below setpoint and 2/2 turbine impulse pressure channels are below setpoint. The P-7 Permissive is active when 2/4 power range neutron flux channels or 1/2 turbine impulse pressure channels are above setpoint. When P-7 is active all previous blocked reactor trips are reinstated. The P-8 Permissive blocks a reactor trip when the plant is below 30 percent of full power on a low reactor coolant flow or a RCP breaker open signal in anyone loop (113 coincidence). The block action (absence of the P-8 Permissive signal) occurs when three out of four neutron flux power range signals are below the 30% setpoint. Thus, below the P-8 setpoint, the reactor will be allowed to operate with one inactive reactor coolant loop and trip will not occur until two loops are indicating low flow. Permissive P-9 shall block a reactor trip following a turbine trip below 35 percent power which is based on the ability of the rod control system and steam dump system to adequately control Tavg on a 50% load rejection (FNP Tech Spec Section B2-8). If2/4 power range nuclear instruments are above 35 percent power, a turbine trip will cause a reactor trip. If 3/4 power range nuclear instruments are below the P-9 setpoint, a turbine trip will not cause a reactor trip. See Table T-3 for a comprehensive list of the protection system blocks. (References 6.4.007, 6.4.011, 6.4.022, 6.7.012) 2.8.2 ESF Permissives

1. P-4 Permissive The P-4 permissive is generated when both the reactor trip breaker and the bypass breaker, which physically bypasses it, are open. Train A of the reactor protection system uses RTA and BYB, and train Buses RTB and BYA. The following are functions ofP-4: a. Causes a turbine trip 2-33 Rev. 10 I FNP Units 1 & 2 REACTOR PROTECTION SYSTEM A-181007 b. Main Feedwater Isolation

-Closes main feedwater regulating valves and feedwater bypass valves if low T avg (554 OF) is also present and requires a manual reset. c. Seals in feedwater isolation signal from safety injection or steam generator high-high water level d. Resets high steam flow setpoint to 40 percent e. Allows operator reset of the safety injection signal after a 60 second time delay This feature ensures that the reactor is tripped and that all emergency core coolant system (ECCS) loads are started before the operator overrides what could be a spurious actuation signal. The block does not prevent the operator from reinitiating safety injection through use of either manual safety injection actuation switch f. Arms steam dumps on a plant trip, defeats the output of the load rejection controller, and places the plant trip controller into control. (References 6.4.007, 6.4.009, 6.7.012 6.7.059, 6.7.060) 2. P-11 Permissive The permissive P-11 allows manual block of low pressurizer pressure safety injection actuation. This permits normal plant cooldown and depressurization. When 2/3 pressurizer pressure bistables sense less than 2000 psig, the low pressurizer pressure safety injection signal may be manually blocked. Placing the train A and train B pressurizer safety injection block switches to BLOCK will now prevent low pressurizer pressure (1850 psig) from initiating safety injection. Each switch will initiate the block function in its respective protective train. If pressure rises above the P-11 setpoint on 2/3 channels, the block automatically clears. If 2/3 channels exceed the P-11 setpoint, and power is available, any shut ECCS accumulator isolation valves will automatically open. In addition, 2/3 pressurizer bistables below the P-ll setpoint blocks the automatic opening of the pressurizer power operated relief valves (PORVs). (References 6.4.007, 6.4.13, 6.7.012, 6.4.017) 2-34 Rev. 10 I

6. 7. speed gain manual rod speed control rods shutdown rods 2. Steam Dump Control A. Impulse unit time constant ofloss ofload interlock channel (pY-447C)

B. Sudden load lou setpoint (C-7) (PB-447A) C. Proportional gain in percent oftota! dump capacity per of 48 stepslminute(l} 62 stepslminute(l) 120 sec. 15% of full load.

Setpoint for full load Im = 577 .ZeF Ja = 567.2°F Loss ofload controller (TY -408]) Plant trip controller (TY -408L) 9.0o/JOF 1) 3.3o/JoF 1) 16. 3 cy'JoF 1) S.O%,oF 1)
For other full load T&VI between 567.2°F and 577.2eF, the proportional gain setpoints should be calculated as follows. (a) Loss of load controller (TY-40SJ).

cy'JGF 100 [(Full Load T &VI -T _Ioed) 12] -Deadband (TY -408 J) (b) Plant trip controller (TY -40SL), cy'cJoF 100 (Full Load T &VI -T_ ... ) -Deadband (TY -408L) There are 8 condenser dump valves. The controlJen should be adjusted. such that the dump capacity is linear with the output ofTY-408J, TY-40aL, QI' PC-464 (bcIGw). That is, the second bank does not begin to modulate opeD \IDtil the first bank has Rev.A7 29 D. E. F. G.

received a signal to modulate full open; etc. The sequence for modulating the valves closed is the reverse of the opening sequence; i.e., the fourth bank to opeD. is the first bank to close, and the third bank starts to close after the fourth bank bas received a signal to close; etc. The first four valves to modulate open are also the first four valves to be tripped open. The last four valves to modulate open are the last four valves to trip open. The two valves in the first bank are designated as the cooldown dwnp valves. The input ranges for modulation (full closed to full open) and the order for modulating the dump valves Ql2m are: First bank (Cooldown valves) Second bank IN/OUT IN/OUT (Fully closed to fully open) 4.0 -8.0 rna => 0.0 -2.5 v=>4 -20 ma (NIC241Y-408S) (NIC241Y-4QS.N) 8.0 -12.0 ma => 2.S -5.0 v=>4 -lDma (NIC241Y-408'I) (NIC241Y-4.0SP)

Third bank 12.0* 16.0 rna => S.O -7.S v=>4 -20 rna (Nl C241Y -408U) (NIC241Y -408Q) Fourthbank 16.0-20.0rna=> 7.5-10.0v=> 4-20rna (NIC241Y-408V) (NIC241Y-408R) Lead time constant vALVES NIN36VSOIA(1)

NIN36VSOU:(1)

N;lN36VSOIC

1) N1N36VSOla

1) NIN36V501B(1)

NIN36VSOIF

1) NIN36V501D(1)

NIN36V50nf l) (1Y-408D) 5 seconds(l) Lag time constant (1Y-408D) 5 seconds(l) Deadband steam dump controller for loss ofload (1Y-408J) 4 o F 1) Deadband steam dump controller for plant trip (1Y-408L) OOF Rev.A7 30 H. I.

High (f wa -T.ae) (TB-408F)

T!YI = 577 .2°F 9.5°F\) 15.l o F t) *Setpomt for full load T!!,I = 567.2°F 7.00pO) First and second bank trip open (High 1) Third and fourth bank trip open (High 2) 10.1 0 1"1) High (f &VI -T DD-Iood) (TB-408J)

Setpomt for full load 15 = 577.2°F T!Ya = 567.2 G F First and second bank trip open (High 1) IS.l o F l) IO.lGF 1) Third and fourth bank trip open (High 2) 30.2 o P<l) 20.:zoF 1)
For other full load Twa between 577.20F and 567.2 0 F, the setpoi.ut.s should be calculated as follows: High CI5 -T!!f) Setpoints (fB-408Fl (a) High 1 (f&VI-Tnoe) valve trip open (TB-408F), of (First and second bank trip open) = [b below -deadband (1Y -408J)]12

+ deadband (IT -408J) (b) High 2 (f &VI -T noe) valve trip open (TB-40SF), OF (Third and fourth bank trip open) == (Full Load T &VI - High CI5 -T 1IO=!ood) Semoints CIB-40SD (a) High 1 (f .... -T..-.) valve trip open (fB-40SJ), of (First and second bank trip open) = [b below -deadband (TY -408L)]J2 + deadband (IT -4OSL) (b) High 2 (T .... -TDO-Iood.) valve trip open (fB-408l). of (Third and fourth bank trip open) = (Full Load Twa -31 Rev.A7 -._-_ ... _--- ." LOAD ReJECTION CONTROLLER (LOW TAVG) 50 4S 40 3$ c-! lao 'Ii t25 20 J 15 J / II J / V 10 / o I o I n << " " 20 T_1EmIr <'MIl* TI9I) o.g F fOF .s Y? (7 T 5::::> 1& F PLAHT TRIP CONTROLLER (LOWTAYG) . / V / ./ 10 / / o / " 20 25 ao T_IEmlr<,ng* T....-) o.g F

Steam Dump Control System (Low T * ..J Rev.A7 32 W 50 -45 40 :15 130 15 lo25 lao J 15 10 40 to * -_._ ....... -._---LOAD REJECTION CONTROLLER (HIGH TAVG) V /. / / / / /v V o . *
a w u U R R ao T_ EmIr (Tavg TNt) Dog f S:-Y7 tP V ,/ $ PUNT TRIP CONTROlLER (HIGH TAVG) /' /' '" V i/' ./' /'" w u ao 25 30 ,J c? d S-'17°F Steam Dump Control System (High Tavel 33 ....... _. __ ....*.*. __ ._._._--_.-

---Rev.A7

J. Header pressure controller (pC-464) set pressure proportional band (based on total condenser dump capacity) reset time constant K. Steam generator relief valve controllers proportional band (valve full stroke) reset time constant set pressure 1005 psig(l) 200 p$i(l) 300 sec.(l) 250 psi(l) 8.5 sec.(l) 1035 psig 3. Pressurizer Pressure Control (Refer to followina figure) A. Pressurizer pressure controller (PC444A) proportional gain reset time constant rate time canstant pressure setpoint, Pref B. Spray valve controllers (pC-444C, PC-444D) proportional gain in percent spray valve liftperpsi 0.5% controller output/psi 600 sec.(l) o sec5 1) 2235 psia (42.S% controller 4o/JOie controller OUtput(l) setpoint where spray is initiated on compensated pressure signal from PC-444A 55% controller OUtput(l)

C. Variable heat controller proportional gain in percent heating power per psi setpoint where proportional heating is full on, on signal from PC-444A 3S% CQIltroller OU1p1K(l) 34 ------------------

Rev.A7 FNP Units 1 & 2 REACTOR PROTECTION SYSTEM A-181007 TABLE T-3 -REACTOR PERMISSIVES AND ESF PERMISSIVES MEASURED COINCIDENCE & MODES OF FSD PERMISSIVE PARAMETERfID SETPOINT *LIOHT STATUS FUNCTION OPERATION SECTION P-4 Reactor Reactor trip and bypass Breakers Open -Either reactor trip breaker Prevents a rapid cooldown of the primary 1,2,3 and 4 2.8.2 trip interlock breakers RTA and BYA and its associated bypass system after a reactor trip Fig. 2, @RTB at;dIWB breaker are open -Trips turbine Sht. 8, -No control board -In coincidence with low T avg' initiates 10,13, indication other than closure of main feedwater reg valves and 15 reactor trip and bypass bypass valves breaker position indication -Prevents opening of main feedwater on the reactor control isolation valves if closed on safety panel injection or SO high-high water level signal -Prevents reactuation of automatic safety injection after safety injection manual reset -Resets high steam flow setpoint to 40% -Allows reset of safety injection signal after a time delay -Defeats the output ofthe Load Rejection Controller -Arms condenser steam dump valves P-6 Intermediate NIS intermediate range 10-10 amps -Allows manual block of Allows power escalation into the 2 below the 2.8.1 range neutron neutron flux channels source range reactor trip Intermediate Power Range by turning P-6 setpoint 2.6.1 flux power NC35D and NC36D on 1/2 intermediate range both MCB Train A and B source range Fig. 2 escalation neutron flux channels block switches to BLOCK above setpoint Sht. 4 permissive > setpoint -Allows manual blocking of source range -Auto reinstates source high neutron flux reactor trip range reactor trip on 212 -Actuating manual block handswitches intermediate range setpoint deenergizes source range instruments neutron flux below -212 intermediate range channels below setpoint setpoint auto reinstates high voltage to -Permissive status light is source range detectors and reinstates lit when 112 channels source range high neutron flux reactor trip > setpoint

Designates Bypass and Permissive Light Box Status B-1 Rev. 5

35. 045Al.05 00lINEWIROICIA 3.8/4.1ININI2/CVRIY Unit 1 was at 30%, and the following conditions occurred:

Control Rods are in Manual.
The Main Turbine was manually tripped. Which one of the following is the initial response of RCS Tavg and RCS Pressure, with no operator actions? Tavg (1) and RCS Pressure (2) (1 ) (2) A'! increases increases B. increases decreases C. decreases increases D. decreases decreases Page: 90 of 200 12/1412009 A -Correct. Steam pressure goes up, causing Tcold to go up. This causes Tavg to go up. This causes a przr insurge which compresses the steam space and RCS pressure goes up. B -Incorrect.

The first part is correct (see A). The second part is incorrect (see A). Plausible, since the insurge is subcooled water from the RCS, but the steam space is compressed which causes pressure to go up. Once the Steam Dumps and/or SG Atmospherics open to reduce SG pressure, the RCS Tavg goes down and the outsurge causes pressure to go down to less than it was initially due to the subcooling of the przr liquid and steam space. C -Incorrect. The first part is incorrect (see A). Plausible, since the Steam Dumps and/or SG Atmospheric relief valves will open to reduce SG pressure and Tavg, but prior to the Steam Dumps and/or SG Atmospherics opening, Tavg will go up. The second part is correct (see A). D -Incorrect. The first part is incorrect (see C). The second part is incorrect (see B). This choice would be the correct response after the initial response. Ran on simulator Laptop (Ie 47) Initial values: Rods in Manual 29% Reactor power 2243 psig 554°F Tripped turbine: Initially: temp, press, went up, power went down. Values 2 minutes later: 16.5% Reactor power stable NI (18% Delta T) 2209 psig 562.5°F (lower than peak of approx. 569°F) Page: 91 of 200 12/1412009 Previous NRC exam history if any: 045A1.05 045 Main Turbine Generator System Al Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MT/G system controls including: (CFR: 41.5 145.5) AI.05 Expected response of primary plant parameters (temperature and pressure) following T/G trip3.8 4.1 Match justification: This question requires knowledge of the expected response (predicting the changes in the parameters) of primary plant Temperature and pressure following the Turbine Trip. Written to preclude steam dump operation due to testing steam dump operation elsewhere in exam. Objective:

15. Describe the operation of the reactor's inherent control systems as they function to re-establish a steady-state condition for the following transients: (OPS52701A

15) a. 10% step load increase b. 50% load rejection

c. 10% ramp increase d. Entry into the power range 18. Determine the final condition of the plant for various transients assuming no operator response (OPS52701AI9).

Page: 92 of 200 12/14/2009

36. 051AA1.04 OOllBANKIROIMEM 2.5/2.5/N/N/2/CVRIY Unit 1 was at 100% power and the following conditions exist:

AOP-B.O, Partial Loss Of Condenser Vacuum, is in progress.
A rapid power reduction per AOP 17.0, Rapid Load Rejection, was completed.
Condenser Vacuum is stable.
FE1, CONT ROD BANK POSITION LO, is in alarm. Which one of the following states: 1) whether or not SS permission is required prior to the Control Rod insertion during the downpower lAW NMP-OS-001, Reactivity Management Program, and 2) the type of Boration required lAW the ARP for FE1? Control Rod Insertion Boration A. SS permission is NOT required.

Emergency boration ONLY. B:t SS permission is NOT required. Normal OR Emergency boration. C. SS permission is required. Emergency boration ONLY. D. SS permission is required. Normal OR Emergency boration. Page: 93 of 200 1211412009 A -Incorrect. Permission not required on insertion. Emergency Boration required until the Rod Bank LO-LO Limit alarm clears. With Rod Bank LO Limit alarm present, a normal boration is required until the LO Limit alarm clears so this is plausible for the student to confuse the the two requirements for normal or emergency boration. B -Correct. Per NMP-OS-001 and ARP-1.6 FE1 & FE2. (See below) C -Incorrect. Permission not required on insertion. Plausible, since it is always required to get SS permission for all positive reactivity additions, and it is expected to get permission when there is time to do so even for negative reactivity additions. However, for responding to a transient to stabilize the plant no permission is required to insert negative reactivity of any type. Emergency Boration is required until the Rod Bank LO-LO Limit alarm clears. With Rod Bank LO Limit alarm present, a normal boration is required until the LO Limit alarm clears so this is plausible for the student to confuse this. D -Incorrect. Permission not required on insertion. Emergency Boration required until the Rod Bank LO-LO Limit alarm clears. With Rod Bank LO Limit alarm present, a normal boration is required until the LO Limit alarm clears so this is plausible for the student to confuse this. NMP-OS-OOl, Reactivity Management Program, Version 13.0 6.3.8.1 During transient conditions that require a rapid reduction in reactor power, operators may take actions to insert negative reactivity that are outside the amounts discussed in the reactivity brief and without SS concurrence. ARP-l.6, FEl Annunciator: CONT ROD BANK POSITION LO, Version 64.0 5. Borate [NORMAL BORATION] the Control Bank "OUT" as necessary using the Boron Addition Nomographs. {CMT 0008900} ARP-1.6, FE2 Annunciator: CONT ROD BANK POSITION LO-LO, Version 64.0 2. Emergency borate the reactor coolant system in accordance with FNP-l-AOP-27.0, EMERGENCY BORA TION. {CMTs 0008555, 0008900} Page: 94 of 200 12/14/2009 Previous NRC exam history if any: 051AA1.04 051 Loss of Condenser Vacuum AAI. Ability to operate and 1 or monitor the following as they apply to the Loss of Condenser Vacuum: (CFR 41.7/45.5/45.6) AAl.04 Rod position .................................................... 2.5* 2.5* Match justification: The question presents a plausible scenario where a rapid power reduction is in progress in accordance with AOP-17. The examinee has to determine the correct procedural guidance given for control rod operation during insertions and withdrawals. The question pertains to whether SS permission is required for insertions and the predicted rod position when Emergency Boration may be terminated. Objective:

6. State the actions that the UO and/or OA TC have the authority to perform in addition to being responsible to the Shift Supervisor (OPS52303HIO).

Page: 95 of 200 12114/2009 07/02/09 06:30:42 FNP-I-ARP-I.6 LOCATION FEl SETPOINT: Variable; 10 Steps Greater than LO-LO Alarm Setpoint. El CONTROD BANK POSITION LO ZLO = ZLO-LO + K4 Where = 10 Steps (6.25 inches) ORIGIN: Rod Insertion Limit Computer PROBABLE CAUSE NOTE:

Zinc Addition System injection will result in a continuous RCS dilution of as much as 1.7 gph, which may result in a reduction in shutdown margin if compensated for by inward rod motion instead of boration.
This annunciator has REFLASH capability.

Reactor Coolant System Boric Acid Concentration too low for Reactor Power Level due to: A. Plant Transient B. Xenon Transient C. Dilution of RCS AUTOMA TIC ACTION NONE OPERATOR ACTION 1. Check indications and determine that actual control bank rod position is at low insertion limit. 1.1 Click on Rod Supervision button on Applications Menu. 1.2 Click on Rod Insertion Limits button. 1.3 Determine if low insertion limit exceeded. IF reactor coolant system dilution is in progress, THEN stop dilution. IF a plant transient is in progress, THEN place the turbine load on "HOLD". Refer to FNP-I-UOP-3.1, POWER OPERATIONS. Borate the Control Bank "OUT" as necessary using the Boron Addition Nomographs. {CMT 0008900} Refer to the Technical Specifications section on Reactivity Control. !f'-f}.j--' C J..-J-f M f.7 ?

References:

A-I77100, Sh. 291; U-26061O; U266647 PLS Document; Technical Specifications DCP 93-1-8587; {CMTs 0008554, 0008887} Page 1 of 1 Version 61.0 07/02/09 06:30:42 SETPOINT: Variable with Reactor Power as measured by .6.T and TAVG. ORIGIN: Rod Insertion Limit Computer PROBABLE CAUSE E2 FNP-I-ARP-I.6 LOCATION FE2 CONTROD BANK POSITION LO-LO NOTE:

Zinc Addition System injection will result in a continuous RCS dilution of as much as 1.7 gph, which may result in a reduction in shutdown margin if compensated for by inward rod motion instead of boration.
This annunciator has REFLASH capability.

1. Reactor Coolant System Boric Acid Concentration too low to ensure Reactor Protection under Accident conditions due to; A. Plant Transient B. Xenon Transient C. Dilution of RCS AUTOMA TIC ACTION NONE OPERA TOR ACTION 1. Check indications and determine that actual control bank rod position is at the low-low insertion limit. 1.1 Click on Rod Supervision button on Applications Menu. A J v:

willi "'\ t-UJ ,.\ /) FNP-I-AOP-27.0, EMERGENCY BORATION . ... r<-{CMTs 0008555, 0008900} o 1 3. IF a plant transient is in progress, '1(\ ?(,.{:-THEN place turbine load on "HOLD". L w fl ! 4. Refer to FNP-I-UOP-3.l, POWER OPERATIONS. A : )hnical Specifications section on Reactivity .p v-X f I" I

References:

A-I77IOO, Sh. 292; U-26061O; U266647 PLS Document; Technical Specifications; DCP 93-1-8587; {CMT 0008887} Page I of 1 Version 61.0 Southern Nuclear Operating Company SOUIHERN'\ Nuclear NMP-OS-001 Management Reactivity Management Program Version 13.0 COMPANY Procedure Page 12 of 29 Erurgy to S""t Your 6.3.2 As a minimum, the Specific Reactivity Management Practices contained in Attachment 2 will be followed. 6.3.3 The SS shall maintain direct supervisory oversight of reactivity manipulations. This {} I yt; the SS will approve each reactivity manipulation, with the exception of c¥ described in step 6.3.8. DUring times where frequent reactivity o ,oJ"-Il manipulations are necessary, the SS can assign a reactivity management SRO to V it? (0 ,7 < "P perform this function while the SS maintains oversight. (:J;J [, 6.3.4 A reactivity brief shall take place at the beginning of each shift in modes 1 and 2. The reactivity brief should include expected reactivity manipulations during the shift needed to maintain current plant conditions or in the case of planned startups, shutdowns or power maneuvers the brief should include a discussion of reactivity changes that would be required to execute these power changes. In addition to this, the reactivity brief should include a discussion of pertinent current core reactivity parameters and any planned work activities that could potentially affect reactivity. The reactivity briefing sheet or OATC turnover sheet shall contain a list of degraded or out of service reactivity manipulation equipment. 6.3.5 When power reduction is necessary, only steam flow adjustments will be effective in reducing and maintaining reactor power below limits. While control rod insertion may appear to provide some immediate relief from high power conditions, the effects are temporary without reducing total steam flow and will only reduce nuclear instrument accuracy due to the resultant cooldown. Turbine load adjustments must be made to reduce and control reactor power, with control rods used primarily to maintain Tave on program during the power reduction. (PWR Only) 6.3.6 Peer checks will be used for reactivity changes, with the exception of conditions described in step 6.3.8. 6.3.7 During some plant operations, one or more of the various indications of reactor power may not be accurate. Therefore, control room operators should always monitor all indications of reactor power and maintain it within licensed limits. 6.3.8 Transient Conditions 6.3.8.1 During transient conditions that require a rapid reduction in reactor power, operators ma take actions to insert ne ative reactivit that are outside the amounts discussed in the reactivity brief an 'l{lthout SS concurrence. The requirement to have another licensed operator peer checR m eac IVlty manipulations under these conditions is also not required since it is unlikely that other licensed operators would be available during the manipulation. The SS shall be briefed as soon as possible on the amount of negative reactivity added (number of steps of rod insertion, amount of boron added (PWR Only), Recirculation Pump speed adjustments (BWR Only), etc.

37. OS4AG2.1.7 OOlINEW/ROIMEM 4.4/4.7/N/N/3/CVRlY A Unit 1 SGFP trip has occurred from 100% power, and the following conditions exist:

AOP-13.0, Condensate And Feedwater Malfunction, is in progress.
The operator is at the step to "Verify automatic operation of the Feedwater Regulating Valves adequate".
SG NR levels are as follows: -1A 34% Rising -1 B 33% Rising -1 C 36% Rising Which one of the following is the correct method of controlling the Main Feed Regulating valves (MFRVs) during this transient lAW AOP-13.0?

Place each MFRV controller in manual at (1) SG NR Level, (2) (1 ) A'I 55% (2) match steam flow and feed flow, and then place the controller back in automatic. B. 55% and then immediately place the controller back in automatic. C. 65% D. 65% Page: 96 of 200 match steam flow and feed flow, and then place the controller back in automatic. and then immediately place the controller back in automatic. 12114/2009 A -Correct. Per step 1.8 and the "D. Operational Concern" note of AOP-13.0. The swell and the response time of the MFRV controller and valve necessitates taking the controller to manual at 55% (before 65% -program level) and matching feed and steam flows prior to taking it back to automatic. This prevents a high high level Turbine trip at 82% level which would occur due to the large Feed Flow Steam flow mismatch if feed flow is not reduced prior to 65%. B -Incorrect. The first part is correct (see A), but the second part is incorrect (see D). C -Incorrect. First part is incorrect, since manual control must be taken at 55% level instead of 65%. Plausible, since doing this at 65% would seem adequate, since that is the level desired to maintain. This could be chosen if the magnitude of the effects of swell and the response time of the controllers is not taken into account. By waiting until 65% to place the controller in manual, the feed flow is high enough that the time to reduce it combined with the expansion of the cooler feed water after getting to the SG can cause excessively high SG levels and a high high SG level trip of the Turbine and SGFPs and a FWIS. The second part is correct (see A). o -Incorrect. The first part is incorrect (see C). The second part is incorrect, since placing in AUTO after taking to manual would not correct the high feed flow and lower the feed flow/steam flow mismatch quickly enough to prevent excessively high SG levels. Plausible, since taking the controller to manual will reset the windup and decrease the controller response time to a level transient, and this is an important part of the procedure guidance reason to go to manual. Also, AUTO control is preferred to manual when adequate for the magnitude of the transient. In smaller SG level transients, going to manual to reset the windup and then allowing AUTO to control the SG level is preferred. FNP-I-AOP-13.0, Condensate And Feedwater Malfunction, Version 29.0 D. Operational Concerns I In the SO level recovery phase, the SO level will start increasing due to the feedwater flow being higher than steam flow and due to swell. If manual action is not taken before the SO reaches normal operating level, the combined effect of swell and additional feed flow may result in SO Hi-Hi Level Turbine and SOFP trip and Feedwater Isolation. Taking manual control and reducing the demand resets the level controller and flow controller integration circuits (i.e. windup) and makes the flow controller output track the associated driver card output. 1.8 Closely monitor steam generator narrow range levels. [ ] WHEN a SO narrow range level recovers to approximately 55%, THEN verifY its main feedwater regulating valve controllers in MANUAL. [ ] Match feed flow with steam flow. [ ] Return feedwater regulating valves to AUTO. Page: 97 of 200 12/1412009 Previous NRC exam history if any: 054AG2.1.7 054 Loss of Main Feedwater 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5 /43.5/45.12/45.13) RO 4.4 SRO 4.7 Match justification: to answer this question correctly, evaluation of plant performance and operational judgement of how to operate the Main Feed Regulating Valves in the given transient condition is required. Instrument interpretation is also included in that with the SG level as low in the Narrow range as they are, MFRV controllers will windup to maximum output by the time the SG level is at normal level of 65%, and this operating characteristic must be taken into account to get the correct answer. Objective:

4. EVALUATE plant conditions and DETERMINE ifany system components need to be operated while performing AOP-IOO, Instrument Malfunction. (OPS-52521Q06).

Page: 98 of200 12114/2009 04/03/09 13 :21: 19 FNP-1-AOP-13.0 CONDENSATE AND FEEDWA TER MALFUNCTION C. Automatic Actions Both MDAFWPs will automatically start on a trip of both SGFPs. Version 29.0 2 The TDAFWP will automatically start at 28% narrow range level in two steam generators. 3 A reactor trip will occur if any SG level decreases to 28% narrow range. 4 A turbine trip and feedwater isolation will occur if any SG level increases to 82% narrow range. 5 A reactor trip will occur if either intermediate range high flux trip bistable (TSLB3-2.1 or TSLB 3-2.2) does not reset before the reactor power is reduced below 10%. 6 Rapidly reducing turbine load may cause the steam dump system to operate in automatic. This will prevent further reduction in total steam flow. Dump operation in steam pressure mode at pot settings other than specified for 1005 psig will affect RCS cooldown rate. D. Operational Concerns In the SG level recovery phaseJ.he SG level will start increasing due to the feedwater flow being higher than steam flow and due to swell. If manual action is not taken before the SG ,.reaches normal operating lev,S(l, the combined effect of swell and additional feed flow may result in SG Hi-Hi Level Turbine and SGFP trip and Feedwater Isolation. Taking manual control an r* and resets the level controller and flow controller cIrcuits I.e. windup) and makes the flow controller output track the associated driver card output. Page 2 of23 04/03/0913:21:19 FNP-I-AOP-13.0 . J. J:;.. JL I.L CONDENSATE AND FEEDWA TER MALFUNCTION Version 29.0 I I I Action/Expected Response Response Not Obtained NOTE:

Steps 1 through 1.3 AND 2 through 2.1 are IMMEDIATE ACTION steps. 1
This procedure steps through probable condensate and feedwater system malfunctions in a systematic diagnostic manner. If the cause of the condensate and feed malfunction is known, THEN the associated procedure section (step) may be implemented immediately.

Single SGFP trip -step 1 Both SGFPs tripped -step 2 SGFP malfunction -step 3 OBSERVE CAUTION prior to step 3 Main feedwater regulating valve malfunction -step 4 Loss of feedwater heater -step 5 SGFP low suction pressure -step 6 Check only one SGFP -RUNNING 1.1 Check generator load GREATER THAN 540MW 1.2 Check rapid turbine load reduction required. 1.3 Check DEHC in OPERA TOR AUTO 1.3.1 Depress the SGFP SETPOINT button on the DEHC keypad 1.3.2 Press the "P8" key 1.3.3 On the SGFP SETPOINT screen verify the following appear: [] A TARGET of"540" MW [] A RATE of"1200" MWlMin 1.3.4 Depress the "GO" pushbutton Proceed to step 2. 1.1 Proceed to step 3 OBSERVE CAUTION prior to step 3. 1.2 IF rapid turbine load reduction NOT required, THEN reduce turbine load using normal DEH controls as required AND proceed to step 3. 1.3 Perform the following a) IF required by generator load

THEN press FAST ACTION AND GV CLOSE pushbuttons
Release the FAST ACTION and GY CLOSE pushbuttons at:::: 730 MWe as indicated on the digital DEHC display b) IF not required by generator load, THEN press the GV CLOSE pushbutton as required to reduce load o Step 1 continued on next page Page Completed Page 3 of23 "i,,,) .,tt;i J: 04/03/09 13 :21: 19 FNP-1-AOP-13.0 CONDENSA TE AND FEEDW A TER MALFUNCTION Version 29.0 I I I Action/Expected Response 1.4 Monitor for correct DEHC system response Response Not Obtained NOTE: A boration of 1 GAL per reduced MW will limit rod insertion and assist in maintaining Delta I. 1.5 Reduce reactor power to match turbine power using control rods and boron. 1.5.1 Verify rod control in MANUAL. 1.5.2 Adjust control rods in MANUAL to reduce reactor power and control RCS TAVG.
Manual Rod Control
Manual boration per FNP-1-S0P-2.3, CHEMICAL AND VOLUME CONTROL SYSTEM REACTOR MAKEUP CONTROL SYSTEM.
Emergency boration per FNP-1-S0P-2.3, CHEMICAL AND VOLUME CONTROL SYSTEM REACTOR MAKEUP CONTROL SYSTEM Figure 6. 1.6 Check proper operation of the steam dumps. 1.7 Verify automatic operation of the feedwater regulating valves adequate.

[] 1ASGFWFLOW FK478 [] 18 SGFWFLOW FK478 [] 1 C SG FW FLOW FK498 1.7 Take manual control of the feedwater regulating valves to control SG level. o Step 1 continued on next page Page Completed Page 4 of23 04/03/09 13 :21: 19 FNP-I-AOP-I3.0 CONDENSA TE AND FEEDW A TER MALFUNCTION Version 29.0 Action/Expected Response 1.8 Closely monitor steam generator narrow range levels. [] WHEN a SG narrow range level recovers to approximately 55%, THEN verifY its main feed water regulating valve controllers in MANUAL. I [] Match feed flow with steam w. * [ ] eturn feed water regulating valves t.£. '""> AUTO. :.....> --1.9 Monitor feedwater flow and steam flow. 1.10 Verify that feedwater and steam flow trend to approximately equal values for the target, turbine load. 1.11 Maintain SG narrow range level approximately 65%. Response Not Obtained 1.8 IF SG narrow range levels NOT maintained greater than 28%, THEN trip the reactor and go to FNP-I-EEP-O, REACTOR TRIP OR SAFETY INJECTION.

CAUTION: The LOSS OF LOAD INTERLOCK C 7 A should not be reset in the event of a failure of PT -44 7 which actuates C-7 A without consultation with the Operations manager. ****************************************************************************************** 1.12 Check LOSS OF LOAD INTERLOCK C-7 A on the BYP & PERMISSIVES panel NOT illuminated.

1. 12 IF C-7 A is to be reset, THEN perform the following 1.12.1 VerifY that all steam dump valves indicate closed. 1.12.2 Verify 0 demand on STM HDR PRESS controller PK 464 and STM DUMP DEMAND TI408 1.12.3 Place STM DUMP INTLK TRAIN A and TRAIN B to OFF RESET 1.12.4 Place STM DUMP MODE SEL TRAINS A-B to RESET and then release to spring return to TA VG. o Step 1 continued on next page Page 5 of23

38. 055EK3.02 OOlIMOD/ROIMEM 4.3/4.6/N/NI2/CVRN ECP-O.O, Loss of All AC Power, directs the operator to:

Dump steam from intact SGs at maximum controllable rate. Which one of the following describes the primary reason for the step which directs dumping steam from intact SGs at maximum controllable rate? A. To minimize potential for SGTR. To minimize RCS inventory loss. C. To maximize TDAFW pump flow. D. To prevent steam voiding in the reactor vessel upper head. A -Incorrect.

This is plausible, since cooling down and reducing the volume of the RCS would reduce RCS pressure, and reducing SG tube dip is desirable, but this is not the reason for the depressurization. B -Correct. This is the Background Document basis for this step: 16.4 in ECB-O.O. (See below). ERG Step Text: The SGs should be depressurized at maximum rate to minimize ReS 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 ReS inventory loss ... C -Incorrect. Plausible, since the ECP-O.O note prior to step 4 does remind of the 2 hour limit on air accumulator supply and UPS power supply, and the heat sink provided by the TDAFW pump is the main source of core cooling with no AC power. Long term, the need to remove decay heat would extend beyond the 2 hours, and a lower SG pressure would allow more water to be pumped to the SG's in the initial 2 hour period. However, the background document requires only SG level of >31 % narrow range on one SG for an adequate heat sink. D -Incorrect. Plausible, since cooling the RCS would eventually cool the vessel head, and without CRDM fans running would be the main cooling for the head. However, the short term as stated in a note in the procedure is that head voiding may be caused by the depressurization. ECP-O.O, Loss Of All Ac Power, Revision 22 ************************************************************************************** CAUTION: The TDAFWP will become unreliable within 2 hours following a loss of all AC power, unless power is restored. This will occur due to a loss of air to the steam supply valves and a loss of control power from the UPS. ************************************************************************************** 4 Verify total AFW flow GREATER THAN 395 gpm. 16.4 Dump steam from intact SGs at 4 Verify proper AFW alignment. Page: 99 of200 12/1412009 maximum controllable rate. FNP-O-EC8-0.0 Section: Procedure Unit 1 ERP Step: 16.4 Unit 2 ERP Step: 16.4 ERG Step No: 16 NOTE-1 ERP StepText: Dump steam from intact SGs at maximum controllable rate. ERG Step Text: 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 ReS 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 FNP-O-EC8-0.0 Section: Procedure Unit 1 ERP Step: 3 Unit 2 ERP Step: 3 ERG Step No: 3 ERP StepText: Verify RCS isolated. ERG Step Text: Check If ReS Is Isolated Purpose: To ensure all RCS outflow paths are isolated Basis: A check for RCS isolation is performed to ensure that RCS inventory loss is minimized. The valves itemized are those in major RCS outflow lines that could contribute to rapid depletion ofRCS inventory .... Following completion of this step, the only ReS inventory leakage path should be the Rep controlled leakage seals .... The secondary depressurization in Step 16 will minimize RCS inventory loss by reducing RCS pressure which will terminate or minimize relief valve flow. For example, reducing RCS pressure to 400 psig would permit the letdown line relief valve to close and would minimize flow through the excess letdown relief valve. Knowledge: Need to minimize RCS inventory depletion during loss of all ac power event to maximize time to core uncovery. Page: 100 of 200 12/1412009 Previous NRC exam history if any: 055EK3.02 055 Station Blackout EK3 Knowledge of the reasons for the following responses as the apply to the Station Blackout: (CFR 41.5 141.1 0/45.6/45.13) EK3.02 Actions contained in EOP for loss of offsite and onsite power .......... 4.3 4.6 Match justification: This question requires knowledge of a response in the EOP (ECP-O.O) which is required to minimize inventory loss during a Station Blackout. Objective:

3. STATE AND EXPLAIN the basis for all Cautions, Notes, and Actions associated with (1) ECP-O.O, Loss of All AC Power; (2) ECP-O.l, Loss of All AC Power Recovery, Without SI Required; (3) ECP-O.2, Loss of All AC Power Recovery, With SI Required. (OPS-52532A03)

Page: 101 of 200 12/1412009 06127/07 16:11:08 Unit 1 ERP Step: 16.4 LOSS OF ALL AC POWER Plant Specific Background Information Section: Procedure Unit 2 ERP Step: 16.4 ERP Step Text: Dump steam from intact SGs at maximum controllable rate. FNP-O-ECB-O.O ERG Step No: 16 NOTE-1 ERG Step Text: The SGs should be depressurized at maximum rate to minimize ReS 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 J:. 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 \7 ? 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 (jJ'" . 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. Knowledge:

References:

SG depressurization should proceed as quickly as possible and should not be limited by the Technical Specification RCS cooldown limit of 100°Flhr. .Justification of Differences: Since the ERG Note contains a directed action it was incorporated into the ERP step for clarification and to enhance procedure flow. 56 of 88 Version: 1.0 06127/07 16:11:08 Unit 1 ERP Step: 3 LOSS OF ALL AC POWER Plant Specific Background Information Section: Procedure Unit 2 ERP Step: 3 FNP-O-ECB-O.O ERG Step No: 3 ERP StepText: Verify RCS isolated. ERG Step Text: Check If RCS Is Isolated Purpose: To ensure all RCS outflow paths are isolated Basis: A check for RCS isolation is performed to ensure that RCS inventory loss is minimized. The valves itemized are those in major RCS outflow lines that could contribute to rapid depletion of RCS inventory. This step is written for plants which utilize air operated valves (AOV s) in the itemized locations. The step structure assumes that the AOV s fail closed on loss of all ac power (i.e., loss of air supply). The operator, therefore, checks that the valves are closed. If any valve is open, the operator should attempt to close the valve. Reasons for a valve remaining open are plant specific, for example the valves may have legitimate or spurious open signals and air pressure could be available due to air receivers or air bottles located in the air supply system. Plants with air receivers may take up to 30 minutes to lose air pressure. If nitrogen bottles are provided for specific valves, such as PORV s, pneumatic pressure may be available for more than 30 minutes. The sequence for checking valves is based on capacity of the outflow lines and potential for RCS inventory loss: 1) The pressurizer PORVs are checked first. Since the turbine_driven AFW pump should be running, the secondary side is removing decay heat and RCS pressure should be under the pressurizer PORV setpoint.

2) The letdown line isolation valves adjacent to the RCS loop are checked second. These valves are normally open and receive a low pressurizer level isolation signal. If these valves, in conjunction with the letdown orifice isolation valves, remain open, a leak path to the pressurizer relief tank (PRT) via the letdown line relief valve may exist. These valves, including the letdown orifice isolation valves, if necessary, should be manually closed as soon as possible to isolate the letdown line and minimize RCS inventory loss prior to automatic isolation on low pressurizer level. Note that isolating the letdown line at the containment penetration will not isolate the letdown relief valve leak path to the PRT. STEP DESCRIPTION TABLE FOR ECA-0.OStep3

3) The excess letdown line isolation valves adjacent to the RCS loop are checked third. These valves are normally closed and do not receive a low pressurizer level isolation signal. If these valves are open a leak path to the PRT via the RCP seal return relief valve may exist. These valves should be closed to isolate the excess letdown line. Note that isolating the seal return line at the containment penetration will not isolate excess letdown inventory loss to the PRT via the seal return relief valve. 4) Any additional plant specific RCS outflow lines should be included.

Following completion of this step, the only RCS inventory leakage path should be the RCP controlled leakage seals. Plants which utilize motor operated valves (MOVs) for letdown or excess letdown isolation will not be able to remotely close these valves to isolate these RCS outflow paths. These plants should isolate these lines at containment. The secondary depressurization in Step 16 will minimize RCS inventory loss by reducing RCS pressure which will terminate or ';-> minimize relief valve flow. For example, reducing RCS pressure to 400 psig would permit the Ce:; ({ t::::.eJ{ letdown line relief valve to close and would minimize flow through the excess letdown relief valve. An alternative which can be evaluated based on plant specific considerations is to dispatch personnel inside containment to manually close the subject isolation valves. 11 of 88 Version: 1.0 06/27/07 16: 11 :08 Knowledge:

References:

LOSS OF ALL AC POWER Plant Specific Background Information Section: Procedure FNP-O-ECB-O.O Need to minimize RCS inventory depletion during loss of all ac power event to maximize time to core uncovery Justification of Differences: Changed to make plant specific. 2 Used action verb "verify" vice "check" since the intent of the step is to ensure the RCS is isolated. 12 of 88 Version: 1.0 FNP-1-ECP-0.0 ,t) ; j LOSS OF ALL AC POWER Revision 22 Step n Action/Expected Response 3.4 Verify all reactor vessel head vent valves -CLOSED. RX VESSEL HEAD VENT OUTER ISO [] Q1B13SV2213A [] Q1B13SV2213B RX VESSEL HEAD VENT INNER ISO [] Q1B13SV2214A [] Q1B13SV2214B Response NOT Obtained ************************************************************************************** CAUTION: The TDAFWP will become unreliable within 2 hours following a loss of all AC power, unless power is restored. This will occur due to a loss of air to the steam supply valves and a loss of control power from the UPS. ************************************************************************************** 4 Verify total AFW flow GREATER THAN 395 gpm. AFW FLOW TO 1A(1B,lC) SG [] FI 3229A [] FI 3229B [] FI 3229C AFW TOTAL FLOW [] FI 3229 4 Verify proper AFW alignment. 4.1 Verify TDAFWP running. TDAFWP STM SUPP FROM 1B (1C) SG [] MLB-4 1-3 lit [] MLB-4 2-3 lit [] MLB-4 3-3 lit TDAFWP SPEED [] SI 3411A > 3900 rpm TDAFWP SPEED CONT [] SIC 3405 adjusted to 100% 4.2 IF TDAFWP NOT running, THEN locally verify TDAFWP TRIP THROTTLE VLV Q1N12MOV3406 open. (100 ft, AUX BLDG TDAFWP room) Step 4 continued on next page. _Page Completed Page 4 of 40 FNP-1-ECP-0.O Step n 'lJ t:i J 1; 1 LOSS OF ALL AC POWER Action/Expected Response Revision 22 Response NOT Obtained ************************************************************************************** CAUTION: Accumulator nitrogen injection into the RCS may result from reduction of SG pressure to less than 100 psig. ************************************************************************************** NOTE: Reduction of intact SGs pressure should continue even if pressurizer level is lost or reactor vessel head voiding j) , .----- 16 Reduce intact SGs pressure to 200 psig. 16.1 Check at least one intact SG narrow range level -GREATER THAN 31% (48%) . 16.1 Perform the following: 16.1.1 Maintain maximum AFW flow to intact SGs until narrow range SG level greater than 31%(48%) in at least one SG. TDAFWP SPEED CONT [] SIC 3405 adjusted to 100% 16.1.2 WHEN narrow range level in at least one intact SG is greater than 31%(48%), THEN perform steps 16.2 through 16.7. 16.1.3 Proceed to step 17. Step 16 continued on next page. ___ Page Completed Page 29 of 40

1. EC-O.OI.l.2-52532A03 003iHL T/IMEM 4.3/4.6/EPE055EK3.021111 055EK3.02 ECP-O.O, Loss of All AC Power, directs the operator to: * "Reduce intact SGs to 200 psig:" Which ONE of the following correctly describes the reason for stopping the SG pressure reduction at 200 psig? A. To prevent losing Pressurizer level. B. To minimize RCS inventory loss out of the RCP seals. C. To prevent steam voiding in the reactor vessel upper head. To prevent injection of SI Accumulator nitrogen into the RCS. Page: 1 of3 10126/2009 EPE055EK3.02 055 Loss of Offsite and Onsite Power (Station Blackout)

EK3 Knowledge of the reasons for the following responses as the apply to the Station Blackout: (CFR 41.5/41.10/45.6/45.13) EK3.02 Actions contained in EOP for loss of offsite and onsite power .. 4.3 4.6 A INCORRECT Plausible, since PRZR level may be lost as stated in ECP-O.O, and is an undesirable condition. Also the note below the caution says the SG pressure reduction should continue if voiding occurs or if Przr level was lost. For this event, depressurization of SGs should be continued even if this does occur so this answer is incorrect. B INCORRECT This is given as the reason why the pressure reduction is being done, not why the pressure reduction is only to 200 psig. This distractor is plausible in that it is closely related to the FINAL PRESSURE to which the SGs are to be depressurized and describes the reason why the depressurization is done, not why it is stopped at 200 pisg. In other procedures, the accumulator MOVs are isolated and and pressure reduction is continued to 100 psig, but with no power, this procedure has pressure maintained at 200 psig. Seal Injection & CCW to thermal Barriers are lost in a LOSP with Loss of all AC power. Seals are a major concern in this procedure. C INCORRECT Plausible, since voiding may occur as stated in ECP-O.O. For this event, depressurization of SGs should be continued even if this does occur so this answer is incorrect. Also the note below the caution in ECP-O.O says the SG pressure reduction should continue if voiding occurs or if Przr level was lost. o CORRECT This is given as the basis.

REFERENCES:

1. ECP-O.O, Loss of All AC Power, Step 28 and NOTES and CAUTION on page 28. CAUTION: Accumulator nitrogen injection into the RCS may result from reduction of SG pressure to less than 100 psig. Finally, the operator should be aware of the limiting low pressure necessary to prevent introduction ofnoncondensibles from the accumulators.

Understanding these considerations, the operator will be able to depressurize and control secondary pressure to minimize ReS inventory loss while minimizing the possibility of introducing nitrogen into the Res and returning the reactor core to a critical condition. Reduce Intact SG Pressure to 200 psig SOs are to be depressurized in the this step to maximize delivery (into the ReS) of the water contained in the SI accumulators while minimizing introducing nitrogen (into the ReS). Introducing nitrogen into the Res could impede natural circulation by collecting in high points of the ReS piping (e.g., SO U-tubes). An ideal gas expansion calculation was used to determine the pressure in the Res following a complete discharge of the contents in the accumulators. This ReS pressure is then correlated to the SO pressure (100 psig), which would be indicated when the accumulators have fully discharged. Page: 20f3 10/2612009 2008 NRC exam Technical

Reference:

ECP-O rev 22 and and the background documents for ECP-O, FNP-O-ECB-O.O Loss of ALL AC Learning Objective: State the basis for all cautions, notes, and actions associated with EEP-3 (OPS52530A03) Comments: -This question matches the KJA in that it asks the applicant to describe the reason for a particular action contained in the EOP for this event. Page: 3 of3 10/2612009

39. 057AG2.4.49 OOlINEW/RO/C/A 4.6/4.4/N/N/3/CVRIY Unit 1 is ramping down for an outage at 2 MW/min. The following conditions occurred:

At 1000:

Reactor power is 25%.
The Reactor Makeup Control System is aligned for repetitive batch borations.
A boration is NOT currently in progress.
LK-112, L TDN TO VCT FLOW, has been adjusted to maintain 45% level in the VCT.
LT-112B and LT-115, VCT LVL, meters both indicate 45%. At 1001 the following occurs:
1 A 120V Vital AC Instrumentation Panel is de-energized due to an electrical fault. Which one of the following is the correct operator response to these conditions?

A. Secure BOTH Reactor Makeup Water Pumps. B. Realign the Reactor Makeup system to AUTO. C'!" Realign LCV-115A, VCT HI LVL DIVERT VLV, to the VCT. D. Increase the ramp rate to control Tavg within the limits of AOP-17, Rapid Load Reduction. Page: 102 of 200 12/1412009 A -Incorrect. The RMW pumps and Boric Acid Transfer pumps don't start unless the RMW control is in Auto when the 1A 120V bus fails. In the conditions given, the system is aligned per repetitive batch borations, which is common during a ramp, and the failure of the 1A 120V bus will not give an auto makeup. Plausible, since if the RMU system was in Auto, OR if an applicant believed that AUTO makeup occurred immediately in the borate mode when the 1A 120V vital panel failed, this choice would be selected. B -Incorrect. Aligning to AUTO is incorrect since it would cause an AUTO makeup to occur regardless of VCT level, and it would not automatically stop. Also,in the repetitive batch Boration alignment, no BAT pump starts and no valves open to cause a boration. Plausible, since confusion between the effects of the 120V vital panel failure in each switch position may exist. In AUTO, the failure causes an Auto makeup to commence. However, in BORATE, an automatic Boration does NOT commence. Failures of other 3 vital instrumentation panels (1 B, 1 C, 1 D) do not cause auto makeups with the control switch in AUTO, so the effects of each of the four 120V Vital panels may be misunderstood. This choice would be selected if an applicant thought that a Boration from the BAT occurred immediately with the switch in BORATE, but that an auto makeup did not occur with the switch in the AUTO position. C -Correct. The LCV115A does divert letdown to the RHT, and will not automatically divert back to the VCT regardless of VCT level per ARP for WD1 :

VCT Hi Lvi Divert Valve -QIE21LCV115A diverts to the RHT if in auto. In addition, an auto makeup will not occur with the switch in BORATE. Prompt action to realign LETDOWN to the VCT must occur to stop the letdown diversion prior to realignment of the RWST to Chg pump suction valves which would cause a significant boration and reactivity event and reactor power transient at the end of life if it occu rred. D -Incorrect.

This is incorrect since Boration from the RWST does not occur immediately requiring this action unless LT-112 is out of service when the 1A 120V vital panel fails. Plausible, since it could occur if LT-112 had also failed or was out of service per the FNP-O-ARP-2.2 WD1: ". IfLT 112 VCT level is out of service, RWST to Chg Pump Suction Valves -Q2E2ILCVI15B & D open." Ran on SIMULATOR from 100% (IC73): IN repetitive boration: No BAT pump starts, the B RMW pump does not start (A RMW pump is normally running all the time), and no valves open (113A, 113B, 114A, & 114B) In AUTO: O/S BAT PUMP STARTS, B RMW PUMP STARTS, 114B RMW TO THE BLENDER modulates open, 113A BORIC ACID TO BLENDER modulates open, 113B MKUP TO CHG PUMP SUCTION HDR opens. Page: 103 of 200 12/14/2009 Previous NRC exam history if any: 057AG2.4.49 057 Loss of Vital AC Electrical Instrument Bus 2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls. (CFR: 41.10/43.2/45.6) RO 4.6 SRO 4.4 Match justification: The Chief examiner was consulted about the difficulty in meeting this k/a with a discriminatory question on the very few immediate actions that are committed to memory. Most "immediate actions" that are performed "without reference to procedures" have been eliminated in recent years. He recommended using actions that must be performed promptly to avoid adverse effects to the plant, whether a procedure directs them or not. This question requires knowledge of the immediate effects of this loss of vital AC Electrical Instrument Bus that would require actions very quickly to mitigate or prevent undesired plant effects. Also, to answer this question correctly, it requires knowledge from memory without a reference being provided such as immediate operator actions. This question fits these criteria, and thus matches the original intent of the KIA. A set of conditions in which a loss of a 120V Vital AC electrical Instrument Bus is given for which prompt actions must be taken to prevent a large boration which would cause a large undesired transient at EOL core conditions. A significant transient, and likely a manual trip requirement, would occur if actions are not taken. Objective:

2. ST ATE AND EXPLAIN any special considerations such as safety hazards and plant condition changes that apply to the 120 Volt AC Distribution System (OPS-52103D04).

Page: 104 of200 12/1412009 07/01109 15:26:52 FNP-0-ARP-2.2 LOCATION WD1 SETPOINT:

1. Battery near exhaustion 107V DC. 2. Inverter output undervoltage 108V AC ORIGIN: 1. Battery near exhaustion X7 Voltage sense board 2. Inverter current limit A3 Ammerter Relay 3. Inverter output undervoltage K3 Relay via X8 voltage sense board 4. Inverter overheating X10 relay board 5. Out of sync X12 relay board 6. Fan failure 7. Bypass source supplying load PROBABLE CAUSE 1. Bypass source carrying load. 2. Inverter out of sync with bypass supply. AUTOMATIC ACTION 1AINV FAULT 1. IF DC input voltage drops to 103 V DC, THEN inverter transfers to bypass source. 2. IF inverter fails, THEN the bypass source should carry the load. 3. An inverter fault when the bypass source is not available resulting in a loss of power to 1A 120 VAC Vital Instrument Panel will be indicated by the following:

A. Source Range Channel 31 will be de-energized. B. Intermediate Range Channel 35 will be de-energized. CAUTION: <)Iltwardrodmotionis blocked by the High Power Rod Stop Bistable being . .. C. NI -41 will be de-energized with associated alarms and indications. D. Annunciators FD3, FD4, DF1 and DK3 will alarm. E. No amperage indication on 1A Inverter ammeter. F. IF RCP breaker indication is lost> 35% power, the reactor will trip. Page 1 of3 Version 26.0 07/01/09 15 :26:52 FNP-0-ARP-2.2 LOCATION WDI OPERA TOR ACTION NOTE: The following controls are affected if 1A 120 V AC Vital Instrument Panel is De-energized:

A TRN SSPS output relay power is lost. ;!O. VCT Hi LvI Divert Valve -Q1E21LCV115A diverts to the RHT ifin auto._ '\; t 1 LTDN Hi Temp Divert Valve -Q1E21TCV143 bypasses the demineralizers. ) t 1A & 1B Reactor makeup water pumps start. M .1A & 1B BAT pumps start. ?> pi I
RMW to Blender -Q1E21FCV114B and Boric Acid to blender -_ ":-l Q1E21FCV113A opens if M/u Control System is in auto. /;) VA l./"'.
LT 11 VCT level is out of service, RWST to Chg Pump Suction Valves -1E21LCV115B

& D open. ji)

Q1E21LCV460 will not close on PZR low level.
Annunciator KG4, TURB TV closed alert, will be in alarm and bistable TSLB2, 14-1 will be lit.
Annunciator KH5, TURB Auto/Stop oil press low, will be in alarm and bistable TSLB2, 13-1 will be lit. 1. IF 1A 120 VAC VITAL INSTRUMENT PANEL is de-energized, THEN immediately perform the following:

A. IF a reactor trip occurs, THEN refer to FNP-1-EEP-0, REACTOR TRIP OR SAFETY INJECTION. B. Attempt to restore power from the bypass source by performing the following:

1. IF the "BYPASS SOURCE A V AILABLE" lamp is illuminated on the inverter, THEN transfer lA INVERTER MANUAL BYPASS SWITCH to the "BYPASS SOURCE TO LOAD" position.

2. IF the "BYPASS SOURCE AVAILABLE" lamp is NOT illuminated on the inverter, THEN perform the following:

Verify IA MCC Energized
Verify Closed Q 1 R17BKRF AF5L Supply to 1 G 208V 1120V REGULATED AC DISTRIBUTION PANEL
Verify closed Breaker #8 IN IG REGULATED AC DISTRIBUTION PANEL
Transfer lA INVERTER MANUAL BYPASS SWITCH to the BYPASS SOURCE TO LOAD position.

Page 2 of3 Version 26.0 07/01109 15:26:52 FNP-0-ARP-2.2 LOCATION WD 1 OPERATOR ACTION cont'd 2. Notify appropriate personnel to determine the cause and correct. NOTE: Per Table 3 of FNP-O-ACP-52.l, Guidelines for Scheduling of On-Line Maintenance, A, B, C, D or F Inverters on bypass source are considered to be unavailable due to being status Al for the Maintenance Rule. This unavailability should be logged for tracking purposes.

2. Refer to Technical Specification 3.8.9 and 3.8.10. 3. IF lA 120 VAC VITAL INSTRUMENT PANEL was de-energized, THEN perform the following when it is re-energized:

A. Verify proper operation of Pressurizer level control and heaters. B. Reset hi flux positive rate trip signal on NI -41 and verify proper operation ofNI-41. C. Verify the following CVCS components are correctly aligned for current plant conditions:

QIE21LCVl15A
QIE21LCV115C

& E

QIE21LCVl15B

& D

QIE21TCV143 D. Verify the following Reactor Makeup Control System components are correctly aligned for current plant conditions:
QIE21FCVI13A
QIE21FCV114B E. Stop unnecessary Reactor Makeup Water and BAT pumps. F. Verify all other MCB controls and indications have returned to normal. 4. Verify control systems outside the control room have returned to normal. 5. WHEN the cause of the fault has been determined AND corrected, THEN return lA Inverter to service. {CMT 0009705} {CMT 0005094} applies to entire annunciator

References:

D-I77024; U-279610; PCN B87-1-2899; D-177218, Sh. 2; D-177214; {CMT 0009705} {CMT 0005094} Page 3 of3 Version 26.0

40. 059A2.04 OOl/MOD/ROICIA 2.9/3.4!N!N13/CYRlY Unit 1 was at 100%, and the following conditions occurred:

The reactor was tripped on simultaneous loss of BOTH SGFPs.
All AFW was subsequently lost.
RCS Bleed and Feed is in progress in accordance with FRP-H.1, Response To Loss Of Secondary Heat Sink.
Core Exit Thermocouples have reached 575°F and are falling.
1A SGFP has been started.
SG Wide Range Levels are: -1A= 8% -1B= 8% -1C= 10% Which ONE of the following describes:

1) the Feedwater flow rate required and 2) the Main Feed System flowpath required for feeding the SGs lAW FRP-H.1? A. 1) Feed ALL SGs at a minimum total flow of 395 gpm. 2) Use the Main Feedwater Regulating Valves. B. 1) Feed ALL SGs at a minimum total flow of 395 gpm. 2) Use the Main Feedwater Regulating Bypass Valves. C. 1) Feed ONE SG at a time at a flow limited to between 20-100 gpm. 2) Use the Main Feedwater Regulating Valves. 1) Feed ONE SG at a time at a flow limited to between 20-100 gpm. 2) Use the Main Feedwater Regulating Bypass Valves. Page: 105 of200 12/1412009 A -Incorrect.

First part incorrect but plausible, since this would be correct if CETCs were less than 550°F, OR if CETCs were their given value and rising instead of falling. Second part is plausible, since the MFRVs are the valves used most often with the SGFPs operating, and are used exclusively on a shutdown until AFW is supplying the SGs. Also, this would be correct per FRP-H.1 Step 7.21 RNO if any of the MFRB valves would not open. B -Incorrect. First part incorrect (see A). Second part correct (see D). C -Incorrect. The first part is correct (see D). Second part is incorrect (see A). 0-Correct. Note prior to FRP-H.1, step 5 states: " IF it is necessary to feed a hot, dry SG(s) [RCS hot leg temperature> 550°F AND SG wide range level <12%{31%}], THEN it (they) should be fed one at a time at a flow rate of 20 gpm to 100 gpm until RCS hot leg temperature falls to less than 550°F. IF bleed and feed is imminent OR bleed and feed is in progress and RCS temperatures are rising, THEN there is no limit on the feed flow rate." FRP-H.1 STEP 7.21 states: "Control feedwater regulating bypass valves to supply main feedwater to intact SGs." FRP-H.1, Revision 26 Previous NRC exam history if any: 059A2.04 059 Main Feedwater (MFW) System A2 Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; 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.l3) A2.04 Feeding a dry S/Q ................................................ 2.9* 3.4* Match justification: The only affects on the MFW system for feeding a hot and dry SG are in a loss of heat sink where the MFW system is required to feed the SG (FRP-H.1). This question requires knowledge of the flow path in the MFW system (impacts of feeding a dry S/G on the MFW system), and the required valve to use and flowrates required in the given condition per the procedure. Objective:

6. EVALUATE plant conditions and DETERMINE if any system components need to be operated while performing (1) FRP-H.1, Response to Loss of Secondary Heat Sink; (2) FRP-H.2, Response to SO Overpressure; (3) FRP-H.3, Response to SO High Level; (4) FRP-H.4, Response to Loss of Normal Steam Release Capabilities; (5) FRP-H.5, Response to SO Low Level. (OPS-52533F06)

Page: 106 of200 12114/2009 8/8/200708:27 FNP-1-FRP-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK Revision 26 Step Action/Expected Response Response NOT Obtained n 4 Monitor CST level. 4.1 [CAl Check CST level greater than 5.3 ft. CST LVL [] LI 4132A [] LI 4132B 4.2 Align makeup to the CST from water treatment plant OR demin water system using FNP-1-S0P-5.0, DEMINERALIZED MAKEUP WATER SYSTEM, as necessary. 4.1 Align AFW pumps suction to SW using FNP-1-S0P-22.0, AUXILIARY FEEDWATER SYSTEM. NOTE:

IF some form of secondary feed flow becomes imminent and normal charging is in service, THEN raising charging flow will reduce the potential subsequent loss of pressurizer level due to cooldown 5 5.1 [ ] [ ] [] shrinkage .
IF it is necessary to feed a hot, dry temperature>

550°F ANQ SG wide range (they) should be fed one at a gpm unti ot eg tempera t an . bleed and feed is imminent OR in temper a tur es ar e tro;' -rate. [CAl Try to establish AFW flow to at least one SG. Verify blowdown from all SGs -ISOLATED. lA(lB,lC) SGBD ISO QIG24HV7614A closed QIG24HV7614B closed QIG24HV7614C closed Step 5 continued on next page. " H-+-6 J ;. )V\.Co-c4 ___ Page Completed Page 4 of 49 T T1\. TTrrl 1 8/8/200708:27 FNP-l-FRP-H.l RESPONSE TO LOSS OF SECONDARY HEAT SINK Revision 26 Step n Action/Expected Response 7.20 Adjust master speed controller to raise feedwater discharge header pressure to 50 psi greater than steam header pressure. FW HDR Response NOT Obtained o f"lt" It cY? ;}-'- 7 PRESS ffi [ ] PI 508 STM y-.t J HDR If-( PRESS t/-0 [ ] PI 464A 7.21 Control feedwater regulating c£ipass)valves to supply main feedwater to intact SGs. Intact SG lA lB lC lA(1B.lC) SG FW BYP FLOW FK [ ] 479 [] 489 [ ] 499 adjusted adjusted adjusted 7.22 WHEN P-12 light lit. THEN perform the following. 7.22.1 Block low steam line pressure SI. STM LINE PRESS SI BLOCK -RESET [] A TRN to BLOCK [] B TRN to BLOCK 7.22.2 Verify blocked indication. BYP & PERMISSIVE STM LINE ISOL. SAFETY INJ. [] TRAIN A BLOCKED light lit [] TRAIN B BLOCKED light lit _Page Completed Page 15 of 49 -ttrf. V ,,!L ,Y'- vJ Locally unlock and control \ main feedwater regulating val ves with handwheel s . (127 7.21 ft. AUX BLDG main steam valve room) Intact SG lA lB lC lA(1B .1C) SG FW FLOW QlC22FCV [] 478 [] 488 [] 498 Key Z-12l Z-120 Z-119

1. FRP-H-52533F03 019IHLTIIC/A 3.7/4.3/w/E05EA2.21111 Page: 1 059A2.04 Given the following:

The reactor was tripped on simultaneous loss of both Steam Generator Feed Pumps.
All AFW was subsequently lost.
RCS Bleed and Feed is in progress in accordance with FRP-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK.
Core Exit Thermocouples have reached 595 of and are still rising.
SGFP "A" has been started.
SG Wide Range Levels are:
A= 8%
B= 8%
C= 14% Which ONE of the following describes the method and rate of establishing feedwater flow at this time? A. Feed rate to A and B SGs is limited to between 20-100 gpm. No limit on the feed rate to C SG. Use the respective Feedwater Control Bypass Valves. B:t There is no limit on the feed rate to any of the SGs. Use the respective Feedwater Control Bypass Valve. C. Feed rate to A and B SGs is limited to between 20-100 gpm. No limit on the feed rate to C SG. Use the Feedwater Control Valves. D. There is no limit on the feed rate to any of the SGs. Use the respective Feedwater Control Valves. See note in FRP-H.1 regarding feeding of hot dry SGs. 2008 Harris NRC Exam Ability to determine and interpret the following as they apply to the (Loss of Secondary Heat Sink) Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.

Technical

Reference:

FRP-H.1, Attachment 1 10126/2009

41. 059G2.1.19 OOI/NEWIRO/C/A 3.9/3.8/N/N/2/CVRIY Unit 1 is at 100%, and the following conditions exist: "-,',

A Computer Alarm for 6A FW HTR Dump valve position is in)
On the IPC, the following indications are observed:

/ 6A FW Heater level indicates 12" and stable. V910A, 6A FW Heater level Dump valve, symbol indicates solid red on the system mimic. V909A, 6A FW Heater level Drain valve, symbol indicates solid green on the system mimic. Which one of the following explains the given indications, and states the isolation signal which will isolate the 5A FW HTR extraction steam valve, V502A, 5A FW HTR ES ISO? 1) 6A FW HTR has a (1) and 2) 5A FW HTR Extraction Steam will isolate on high (2) (1 ) (2) A. tube leak Level B. tube leak Pressure C'!" failed open dump valve controller Level O. failed open dump valve controller Pressure A -Incorrect. The first part is incorrect, since the FW HTR is actually empty. Below 14" tubes are uncovered and the htr is actually empty with steam blowing by to the HOT. Plausible, since indication is at 12" and the level indication is steady. A tube leak would cause this indication in other heaters (except for 6A & B) except for the drain valve being closed. If a tube leak were causing this indication, the drain would be open also. The second part is correct. Plausible even when combined with the incorrect first part, since it is correct for a HTR malfunction which causes steam flow through the 6A FW Htr to the HOT, but not for high liquid flow to the HOT. Often with a transient in the FW HTR strings, other heaters are affected. Confusion may exist as exactly how 6A FW HTR will affect the HOT and 5A FW HTR to cause the extraction valve to close. B -Incorrect. The first part is incorrect (see A). The second part is incorrect. Plausible, since the number 6 heater can cause a high level in the HOT which would cause a high pressure in the 5A FW heater, however, there is no high pressure isolation in the 5A HTR, only a high level. Also, the 5A heater does not normally have any level, and the incorrect assumption could be made that it does not have a high level isolation. The 6A FW heater has a dIp isolation Signal, and confusion could exist as to the 5A isolation signals. Page: 107 of200 12114/2009 C -Correct. Cautions in ARP-1.1 0, for HTR H I LEVEL and in SOP-20 state that if the 6A HTR level drops below 14" (18-19" is normal level controlled by the drain valve), then the tubes will be uncovered and the tank will empty. Even when empty, the tank will indicate 12". The dump being open below the normal level of 18-19" and the drain being closed is indication that the dump valve controller is failed to demand full open, or at least controlling at too low of a level. The same cautions state that with the 6A FW HTR empty, steam will blow by to the HDT, pressurize the HDT AND the 5A FW HTR, and extraction steam will isolate on high 5A FW HTR level. D -Incorrect. The first part is correct (see C). The second part is incorrect (see 8). Plausible, since it may be understood that below 14" the heater empties and blows steam to the HDT instead of subcooled liquid, but confusion may still exist as to the relationship between the HDT pressure going up, causing the 5A FW HTR pressure to go up, and the 5A FW HTR extraction steam isolating on high level. Also, the 5A heater does not normally have a level, and the incorrect assumption could be made that it does not have a high level isolation either, but could have a pressure isolation. ARP-1.10, KC4, FW HTR ORDRN TK LVL HI, Version 64.0 AUTOMATIC ACTION If level is not stabilized, extraction steam to the HP and LP heaters will automatically isolate. See Table 1 for information next page. CAUTION: DO NOT let #6 FW HTR level trend below 14". If the tubes are uncovered level will indicate 12", but steam is actually blowing by and pressurizing the HDT. This results in level increase in the #5 FW HTR due to its [IN]ability to drain to the HDT. On high level, the #5 FW HTR extraction steam will close. (AI 2008202332) FNP-1-S0P-20.0, FEED WATER HEATER EXTRACTION, VENT AND DRAIN SYSTEM, Version 53.0 3.7 If the tubes in #6 FW HTR are uncovered level will indicate 12", but steam is actually blowing by and pressurizing the HDT. This results in a level increase in the #5 FW HTR due to its inability to drain to the HDT. On high level, the #5 FW HTR extraction steam will close. (AI 2008202330) Page: 108 of200 12/1412009 Previous NRC exam history if any: 059G2.1.19 059 Main Feedwater System 2.1.19 Ability to use plant computers to evaluate system or component status. (CFR: 41.10/45.12) RO 3.9 SRO 3.8 Match justification: This question requires the evaluation of plant computer points for some parameters for a FW system component (6A FW Heater) to determine component status of two FW system components: 6A FW HTR actual level and 5A FW HTR extraction valve resulting status. Objective:

6. DEFINE AND EVALUATE the operational implications of normal/abnormal plant or equipment conditions associated with the safe operation of the Main and Reheat Steam System components and equipment, to include the following (OPS-40201A07):

Normal control methods
Abnormal and Emergency Control Methods
Automatic actuation including setpoint (example SI, Phase A, Phase B, MSLIAS, LOSP, SG level)
Protective isolations such as high flow, low pressure, low level including setpoint
Protective interlocks
Actions needed to mitigate the consequence of the abnormality Page: 109 of 200 12/1412009 07/02/0906:30:47 FNP-I-ARP-I.IO LOCATION KC4 SETPOINT:

1. Heater Drain Tanks: 4.25 inches above center of tank. FWHTR ORDRNTK level. HEATER: 2. L.P. Heaters: 3.0 inches above normal liquid 3. H.P. Heaters: 3.0 inches above normal liquid level. APPROXIMATE MCB LEVEL IA&B l3" HTRDRNTK:

A 51.75" 2A&B 9" B 51.75" 3A&B l3" 4A&B 8" 6A&B 21" LVL HI ORIGIN: The following Level Switches: a) lAHDTNlN26LI502 e) 4AHeaterNIN21LI538 I) b) lA Heater NIN21LI526 f) 6A Heater NIN21LI542 j) c) 2A Heater NIN21LI530 g) lB HDT NIN21LI504 k) d) 3A Heater NIN21LI534 h) lB Heater NIN21LI528

1) PROBABLE CAUSE 1. Heater tube leak or rupture. 2B Heater NIN21Ll532 3B Heater NIN21Ll536 4 B Heater N 1 N21 Ll540 6B Heater N1N21LI544

2. Malfunction of a Level Controller on a Low Pressure or High Pressure Heater or Heater Drain Tank. AUTOMATIC ACTION If level is not stabilized, extraction steam to the HP and LP heaters will automatically isolate. See Table 1 for information next page. CAUTION:.

DO NOT let #6 FWHTRleveltrendbelow 14". If the tubes are u OPERA TOR ACTION Determine which heater or drain tank level is high. Verify that the dump to condenser valve of the affected heater or drain tank is open. IF a heater tube leak OR rupture is indicated, THEN isolate the affected heater. IF a feedwater heater malfunction is indicated, THEN go to FNP-I-AOP-l3.0, CONDENSATE AND FEEDWATER MALFUNCTION Refer to FNP-I-SOP-2I.0 for limitations on Turbine operation with one or more Feedwater Heaters isolated. Monitor Feedwater Heater and Drain Tank Levels to ensure that they are returning to normal. Page 1 of2 Version 62.0 07/02/0906:30:47 FNP-I-ARP-1.l0

7. Notify appropriate personnel to determine and correct the cause of the alarm. Page 2 of2 Version 62.0 07/02/0906:30:47 FNP-I-ARP-l.10 TABLE 1 Heate Setpoint for Extraction Steam Extraction Stm Level Switch Supply Breaker r MOV Isolation MOV lA 18" above HLL NIN39LS502A NIN35V517A-N NIRI7BKRHBBB3

-18" above HLL NIN39LS502B NIN35V518A-N NIR17BKRHBBB4 ';' ,/ . .eN'//; >; ,;; <" ;/ .. *;*7 7 1<;;: IB 18" above HLL NIN39LS502A NIN35V517B-N NIR17BKRHBBB3 III 18" above HLL NIN39LS502B NIN35V518B-N NIR17BKRHBBB4

";;' .. 2A 18" above HLL NIN39LS505A NIN35V519A-N NIR17BKRHBBC4 2B 18" above HLL NIN39LS505B NIN35V519B-N NIR17BKRHBBC5 3A 18 y,." above NIN39LS508A NIN35V506A-N NIR17BKRHAAA4 HLL 3B 18 y,." above NIN39LS508B NIN35V506B-N NIR17BKRHAAA5 HLL 4A 14" above HLL NIN39LS510A NIN35V507A-N NIR17BKRHBBA6 4B 14" above HLL NIN39LS510B NIN35V507B-N NIR17BKRHBBB2 18" above HLL NIN39LS512A NIN35V502A-N NIR17BKRHAAB2 .f1C-5B 18" above HLL NIN39LS512B NIN35V502B-N NIR17BRKHAAB3 18" above HLL NIN39LS515A OR OR 6A closes @ 0-2/+0 NIN35PDS547A ORB NIN35V503A-N NIR17BRKHAAA6 opens@ 6 +/- 2 .----18" above HLL NIN39LS515B OR OR 6B NIN35V503B-N NIR17BKRHAAB5 closes @ 0-2/+0 NIN35PDS547A OR B \ opens@6+/-2 A A-I77100/459; A-170750/88&89; D-I72570 thru D-I72577; D-I7257811

&2: -D-I7257911&2; B-175968; B-170058/97&98; A-170256; A-170257 Page 3 of2 Version 62.0

42. 061G2.2.37 OOllFNP BANKIROICIA 3.6/4.61N1N13/CVRIY At 1000 the following plant conditions exist on Unit 1:

A TECH SPEC required shutdown was in progress due to BOTH 1A and 1 B SW pumps inoperable and unavailable (not running).
1 C SW pump is aligned to B Train. At 1005 the following events occur:
A seismic event caused a loss of BOTH SGFPs, a leak in the A Train SW header and a tear in the CST at the bottom of the tank.
CST level is at 5 ft. and decreasing.

Which one of the following describes the purpose of the actions directed by SOP-22.0, Auxiliary Feedwater System? To establish availability of ______ _ A. 1A and 1 B MDAFW pumps with SW valve alignments made from the main control room ONLY. B. the TDAFW and the 1 B MDAFW pumps with SW valve alignments made from the main control room ONLY. C. ALL AFW pumps with SW valve alignments made from in the plant AND from the main control room. Dy the TDAFW pump with SW valve alignments made from in the plant AND from the main control room, and the 1 B MDAFW pump with SW valve alignments made from the main control room ONLY. Page: 110 of 200 12/1412009 A -Incorrect. Correct for B MDAFW pump only, and incorrect for A MDAFW pump. B has a source from B train SW with MCR valve alignments only. Even though B MDAFW pump has a suction source from B train SW, A MDAFW pump has no procedurally allowed suction source from A train SW per SOP-22. However, the system would allow cross connecting trains to supply both A and/or B MDAFW from B train SW. Plausible, since in this emergency, it could be incorrectly assumed that maximum flexibility would be written into the procedures to allow this option to cool the SGs and Core. B -Incorrect. Correct for B MDAFW Pump. TDAFW has a source of suction from A train only aligned to enable supplying with only MCR MOV operations. Plausible, since the TDAFW suction can be supplied from B train AFW with manual valve alignments outside of the Control Room. C -Incorrect. Correct for B MDAFW and the TDAFW Pumps. Incorrect for A MDAFW pump. Plausible, since the system is versatile enough to allow cross connecting trains to supply both A and/or B MDAFW from B train SW. In this emergency, it could be incorrectly assumed that maximum flexibility would be written into the procedures to allow this option to cool the SGs and Core. o -Correct. Procedurally, B MDAFW (with MCR valve alignments only) and TDAFW pump (with in-plant and MCR valve alignments required) are both able to use B train SW as an auxiliary suction source. SOP-22, AUXILIARY FEEDWATER SYSTEM, Version 64.0 Page: III of 200 12/14/2009 Previous NRC exam history if any: 061G2.2.37 061 Auxiliary / Emergency Feedwater System 2.2.37 Ability to determine operability and/or availability of safety related equipment. (CFR: 41.7/43.5/45.12) RO 3.6 SRO 4.6 Match justification: Determining availability is more on the RO level than determining operability (other than determinations of low discriminatory value such as pump trips on overcurrent, no power to start pumps, etc.). This question requires knowledge of which of the AFW pumps are available from only one train of their alternate suction source of Service Water when their primary suction source, the CST, is not available. It also requires knowledge of how the pumps will become available from their alternate suction source procedurally and by location of the valve manipulations. Objective:

7. DEFINE AND EVALUATE the operational implications of normal/abnormal plant or equipment conditions associated with the safe operation of AFW System components and equipment to include the following (OPS-40201D07):

Page: 112 of 200

Normal Control Methods
Abnormal and Emergency Control Methods
Automatic actuation including setpoints (examples

-SI, Phase A, Phase B, MSLIAS, LOSP or SG level)

Actions needed to mitigate the consequence of the abnormality 12/14/2009 07/02/09 06:34: 16 FNP-I-S0P-22.0 4.7 Aligning Service Water to the AFW System. CAUTION: Service water does not meet secondary makeup specifications and should only be used when required by emergency conditions.

NOTES:

Refer to Tech. Specs. section 3.7.6 Condition A.
2 BOP keys from the Shift Support Supervisor's office will be required in the following steps. 4.7.1 Notify Shift Chemist that SW will be added to the SO's. NOTE: FNP-I-SOP-24.0, SERVICE WATER SYSTEM, requires starting/stopping SW pumps as necessary to maintain pressure in each header greater than 70 psig but less than 130 psig as indicated on PI-3001AA and PI-3001BA or by adjusting CCW HX DISCH FCV HIC 3009 A(B, C). 4.7.2 Verify service water is in operation per FNP-I-S0P-24.0, SERVICE WATER SYSTEM maintaining proper SW pressure.

4.7.3 Open MDAFWP SW SUPP: (BOP key operated switches)

QIN23MOV3209A It {-rV\
QIN23MOV3209B 4.7.4 Open: (BOP)
MDAFWP SW SUPP QIN23MOV3210A
MDAFWP SW SUPP QIN23MOV3210B
TDAFWP SW SUPP QIN23MOV3216. Version 61.0 07/02/09 06:34: 16 FNP-I-S0P-22.0 4.7.4.1 IF necessary to align TDAFWP suction from B Train service water, THEN perform the following.

NOTE: During the following two steps service water trains will be temporarily cross-connected. 4.8 4.7.5 4.7.4.1.1 Unlock and water to TDAFWP suction:

QIN23V015D (in IB MDAFWP Room) *-* QIN23V015C (above MDAFWP Room) 4.7.4.1.2 Unlock and close{A TralDlservice water to TDAFWP suction:
Q1N23V015B (above MDAFWP Rooms) +-* Q1N23V015A (in 1A MDAFWP IF required, THEN place AFW system in operation per 4.1 or a..!) . rna IA. (.A 4.3 of thIS SOP. t\ -J(! ,//vtt 1 \)e ;; Condensate Storage Tank Feed and Bleed. 4.8.1 Verify CST LVL LI 4005B 2: 12.5 ft and CST LVL LI 4132B is reading maximum indication.

4.8.2 Verify with Chemistry that water is within specifications to be drained to the yard drains. 4.8.3 Remove blind flange and throttle open CST drain valve Q1P11V508. 4.8.4 Verify CST L VL LI 4005B 2: 12.5 ft and CST L VL LI 4132B is reading its maximum indication at least once every 30 minutes while feed and bleed operation is in progress. 4.8.5 IF either level indication criteria is NOT met, THEN close CST drain valve (Q1P11V508) and restore CST level per FNP-I-S0P-21.0, CONDENSATE AND FEEDWATER SYSTEM. 4.8.6 WHEN feed and bleed operation is complete, THEN close CST drain valve (QIP11V508) and replace blind flange. 4.8.7 Verify CST LVL LI 4005B 2: 12.5 FT and CST LVL LI 4132B is reading its maximum indication. Version 61.0 Date: 10r\:2009 A B c D E , 1 I 2 3 4 4-;fL-P /0 bd 'r ....... / /;C' I I I \ I 8 HBC-'L-l--I' , 1--)0-175003 SH.2(C-3)<III 2 ,I_""M M ... , SERVICE WATER _" . ___ I 0-175007 (C-l0) 2" HBD-46 0-175007 (E-l0) 2" HBO-46 0-175007 (H-l0) 2" HBD-46 <D /;-1 ....... , f I i /./""--.., 55 /// -..., N1N23V0137

t: 0;-I e. " :i en o o o It) I'--I o I I I I I I I I I I I I I I I I I I L_.f\._' __ .f'\._, '" " I U <Xl :t: : ' I '" I I i I I u I I I I 5 Lii"L (PT RED. /rI\ 6 Time:

36 AM 7 f) 3/4" DBC-2 V064E V064F V043A ""'/--'-----1

(, .3/4" DR. 4" DBC-2 LO V019A 1 I I I LC 1 V009A v I (J ill o ':.,-V001A V025A V044A 1" DR. V061A V025E rFO\J I o <Xl o :.,. I ur' I 'ti'bHt L= __ , __ ]"-I en (J I fi I I rJ----L----- ___

t:1 I I L----l NV004 1'--' v I V060 > , <D ...... -"""0 "'> LO V001B V045A LO V017A PUMP AUTO-START (lEE NOTE 8 -b'C /' Vb59Ai

/ o /O:T:\:7 : @:-:--fFo\ : s: I ,:t:1 i @; v 1,0 /1 M M :D: I ),,J, I V025B V025D I 1-_-' LO Lcfvo09B V019B.LO 1"8/ V053 V017E ']_-1 _ FD I : /, V057 8618 I-t:J : L_ l*DR. ." _M __ ... 1 V002B----..,

____ -fi3: 2" DBC-4 I I I 862B r ,0 -;'::]; / : P::S:'J '7\ Title: C:\Reference Disk\Exam Reference Disk\Drawings\D175007 -0001.cal Date: 10r E F G H J 0:> ..,. I 2" HBO-46 0-175007 (C-10) u lD :r: I I I 0-175007 (['::15)) 2" HBO-46 I I I I I I I 0-175007 (i-t'::jO) 2" HBO-46 L_J'\._' __ J'\._, : I " LI '\ El. (' I I I I I I I I r :r: I:r: I 7 IN ro I ro : i I i STOIViGCTAI'll< Q1Pl1T001 500,000 GAL. ,1, X LO 'T' I I I I I I T <0 ..,. I Cl CD :r: NV003 lO NV012 1" DR. 3/4" X 1/2" RED. I : 40 I V062A V062B L_ rC<<J-J,rt-:ll I VUIOt ;--1J tr _ _ V035 V034 I -, ,'", @ C'" co . , "" '" ... ,' " "",-" >00--' * , ..... , > 0-175000 SH.2(A-2l CHEMICAL 1 @ -BECHTEL FI 3403A SEE D-170117 SH.2 (F-4) FOR CONTINUATION 2 INJECTION SYS. 3 4 Time: '\53:08 AM J .,. 'f-4" HBD-47 4" HBD-47 6" HBO-46 2" OBC-4 _ 3" OBC-4 3" X 2" RED. a5 Cl N '----" U Cl U lCI'V009B V040 5 TE 293L 6 V003 V025B V025D V053 FD V057 861 1" DR. voss t / ,/ " 7 Title: C:\Reference Disk\Exam Reference Disk\Drawings\D175007 -0001.cal

43. 062Al.Ol OOllNEW/ROICIA 3.4/3.8ININI4/CVRlVER 5 EDITORIAL Unit 1 has had a LOSP followed by a SBLOCA, and the following conditions exist:

The 1-2A DG is tagged out.
The 1 B DG is tripped.
The 1 C DG is supplying power to the A Train busses.
The load on the 1 C DG is 2.860 MW. Which one of the following describes whether or not the 2000 amp hour rating on the 1 C DG will be exceeded if the 1 B PRZR HTR GROUP BACKUP is energized?

lE the PRZR HTR GROUP BACKUP is energized, THEN the 1 C DG 2000 hour load rating (1) be exceeded, and energizing the 1 B PRZR HTR GROUP BACKUP (2) allowed lAW EEP-1, Loss of Reactor or Secondary Coolant. (1 ) (2) A. will is will is NOT C. will NOT is D. will NOT is NOT Note: in this alignment the 1 C DG has been manually aligned to the 1 F bus. A -Incorrect. The first part is correct (see B). The Second part is incorrect (see B). Plausible since the procedure does state that in an emergency, the design of the electrical system has determined that a slight overload may exist after a LOSP and a LOCA. This is acceptable as long as the 2000 hour rating is not exceeded. Also, the Basis of TS 3.8.3 allows the 2000 hour limit to be exceeded for up to 300 hour per year but EEP-1 forbits exceeding the 2000 hour limit. Confusion may exist as to which one or if both the continuous and/or the 2000 ratings may be exceeded for a period of time. Also, the continuous load limit and the 2000 hour rating values may not be remembered properly. However, the procedure states that MANUAL loading above EITHER the continuous or 2000 rating is not allowed. Confusion could exist as to what the 2000 hour load allows, i.e. it does not allow overloading above the limit for any period of time as the continuous load limit does . . B -Correct. The continuous'16a'd limit is and the 2000 hour load limit is .. ,,-<: . .1-10,,9 MW. > continuous load limit. EEP-1 APP 4, Step 1 caution states Do NOT manually load diesel generators above 2000 hr. load limit. Per EEP-1, Att. 4: " ... continuous load rating limit (i.e. 2.85 MW for small DGs, 4.075 MW for large DGs). Under these circumstances, diesel generator loading may be raised not to exceed the 2000 hour load rating limit (i.e. 3.1 MW for small DGs, 4.353 MW for large DGs ... ). Page: 113 of 200 12/14/2009 C -Incorrect. The first part is incorrect. Plausible, since the continuous load limit and the 2000 hour rating may not be remembered properly, and/or the load in MW of the przr heaters may be confused with other smaller loads. The large DGs have a higher load limit of 4.075 & 4.353 MW for continuous and 2000 hr rating respectively. The second part is incorrect (see A). 0-Incorrect. The first part is incorrect (see C). The second part is correct (see 8). Plausible, since the continuous load liming is already exceeded, and manual loading above the continuous load is not allowed, even though automatic loading above the continuous load limit is allowed in an emergency. FSD, Diesel Generator System 3.1.6 Interface Requirements The only time during operation (other than design basis accidents) that the diesel is intentionally loaded above its continuous rating is during Technical Specification surveillance testing when the diesels are loaded to their 2000h ratings (4353 KW for the large diesels and 3100 KW for the small diesels). APPENDIXB STATIC LOADING OF THE DIESEL GENERATORS B.

2.0 INTRODUCTION

During some design basic events, diesel generator 1 C is loaded above its continuous rating by less than 5%. However, this calculated loading above the continuous rating is acceptable since the diesel loading still meets the criteria contained in Position C.2 of Safety Guide 9 (Reference 6.7.028). G.4.3 Potential Diesel Generator Overload The potential exists for DG overload if the LOSP is followed by a LOCA after step 2 of the LOSP sequencer has been energized. In those cases, the DG will be loaded with the Reactor Cavity Cooling Fan (13 Kw) and the CRDM fan (84 Kw) in addition to the ESS loads, and the operator may have to remove selected loads if the DG is loaded above its rated capacity. This situation does not constitute a concern given the existing guidance in the plant procedures which provides the operator with guidance for reducing the DG loading if it is above rated capacity. EEP-1 LOSS OF REACTOR OR SECONDARY COOLANT Revision 29 ATTACHMENT 4 VERIFYING 4160 V BUSSES ENERGIZED 1 Verify 4160 V busses energized.

CAUTION: IF a DG is already operating above its continuous load rating, THEN additional manual loads should not be added. Unanticipated plant emergency conditions may dictate the need to load the emergency diesel generators above the continuous load rating limit (i.e. 2.85 MW for small DGs, 4.075 MW for large DGs). Under these circumstances, diesel generator loading may be raised not to exceed the 2000 hour load rating limit (Le. 3.1 MW for small DGs, 4.353 MW for large DGs). Diesel loading should be reduced within the diesel generator continuous load rating limit as soon as plant conditions allow. ************************************************************************************** Page: 114 of200 12/1412009

CAUTION: To prevent diesel generator overloading, at least 0.3 MW of diesel generator capacity must be available prior to energizing a group of pressurizer heaters. ************************************************************************************** 1.7.4 RNO Energize pressurizer heater group 1 B as required. Previous NRC exam history if any: 062A1.01 062 A.C. Electrical Distribution Al Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including: (CFR: 4l.5 145.5) Al.Ol Significance of DIG load limits .................................... 3.4 3.8 Match justification: In this question parameters are provided which must be evaluated to predict if the DG will be overloaded if a load is manually started. The size of the load in MWs must be known and the load limit must be know to predict if the load may be started and if it will exceed the design limits. The significance of the load limits must be understood, since the continuous limit may be exceeded in an emergency without expected damage to the DG, but the 2000 hour load limit may not be exceeded for any reason. Objective:

2. RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the components associated with the Diesel Generator and Auxiliaries System, to include the following (OPS-40 1 02C02): a. PC2 Diesel, including capacity b. FM Diesel, including capacity Page: 115 of 200 12/1412009

.1 8/8/2007 08: 19 FNP-1-EEP-1 LOSS OF REACTOR OR SECONDARY COOLANT Revision 29 Step n Action/Expected Response ATTACHMENT 4 1.5 Verify BKR DG02 (lG 4160 V bus tie to 1L 4160 V bus) -CLOSED. l.6 Verify all RCP busses -ENERGIZED. [ 1 1A 4160 V bus [ 1 1B 4160 V bus [ 1 1C 4160 V bus l.7 Check IE 4160 V bus -ENERGIZED. Response NOT Obtained 1.5 IF diesel generator cooling NOT supplied. THEN secure 1B diesel generator using ATTACHMENT

5. SECURING A DIESEL GENERATOR WITH A SAFETY INJECTION SIGNAL PRESENT. 1.7 Establish power to 1C 600 V LC emergency section loads. 1.7.1 Place handswitch for pressurizer heater group 1B in OFF. 1.7.2 Open BKR ECOS-1. 1.7.3 Close BKRs EE07-1 and EClO-l. **************************************************************************************

CAUTION: To prevent diesel generator overloading. at least 0.3 MW of diesel generator capacity must be available prior to energizing a group of pressurizer heaters. * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * -3 / *bO w 1.S Check 1D 4160 V bus -ENERGIZED. 1.9 IF 1D 4160 V bus energized. THEN return to PROCEDURE STEPS. step 13. 1.7.4 Energize pressurizer heater group 1B as required. 1.S Proceed to step 1.10. Step 1 continued on next page. _Page Completed Page 2 of 6 8/8/200708:19 FNP-1-EEP-l LOSS OF REACTOR OR SECONDARY COOLANT Revision 29 Step n 1 Action/Expected Response Response NOT Obtained ATTACHMENT 4 VERIFYING 4160 V BUSSES ENERGIZED Verify 4160 V busses energized. 1/ 4--C-?-'A-d-i6--T+-G..'+;P '?-

CAUTION: IF a DG is already operating above its continuous load rating. THEN additional manual loads should not be added. Unanticipated plant emergency conditions may dictate the need to load the emergency diesel generators above the continuous load rating limit (i.e. 2.85 MW for small DGs. 4.075 MW for large DGs). Under ____ __ 10 be raised not to for smal within the diesel s on as plant conditions >11/ C,OH\W> 3) 11\ * * * * * * * * * * * * * * * * * * * * * * * * *

* * * * * * * * * * * * * * * * * * *
*.:t;* * * * * * * * * * *
* * * * * * * * * * * * * * * * * * * * * * * * *
A -I-f3. I 7

$> L C 4: i> I i' -C V j \4...C.q--f" NOTE: Plant conditions may dictate establishment of contingency electrical lineups. FNP-1-AOP-5.1. CONTINGENCY ELECTRICAL ALIGNMENTS provides gUidance for establishing those lineups. 1.1 Check offsite power -AVAILABLE. 1.2 Check BKR DF01 (lA startup transformer to IF 4160 V bus) -CLOSED. 1.3 Verify BKR DF02 (IF 4160 V bus tie to lK 4160 V bus) -CLOSED. 1.4 Check BKR DG15 (lB startup transformer to 1G 4160 V bus) -CLOSED. 1.1 Request Shift Manager coordinate efforts to restore offsite power. 1.2 Verify IF 4160 V bus energized by 1-2A or 1C diesel generator. 1.3 IF diesel generator cooling NOT supplied from Unit 2. THEN secure 1-2A and/or 1C diesel generator using ATTACHMENT

5. SECURING A DIESEL GENERATOR WITH A SAFETY INJECTION SIGNAL PRESENT. 1.4 Verify 1G 4160 V bus energized by 1B diesel generator.

Step 1 continued on next page. ___ Page Completed Page 1 of 6

44. 063A3.01 OOI/FNP BANKIROICIA 2.7/3.11Y 2007/N/4/CVRlVER 5 EDITORIAL Unit 1 is at 100% power with the following conditions:

1A Battery Charger is on service.
EM personnel are doing preventative maintenance on the 1A battery. The following indications and alarms are received:
The UNIT 1 AUX BLDG DC BUS -A TRN GROUND DET white light comes on momentarily and then goes OFF.
WC3, 1A 125V DC BUS BATT BKR 72-LA05 TRIPPED
WC2, 1A 125V DC BUS UV OR GND alarms and clears. Which ONE of the following describes the status of the indications on the EPB for the 1A DC BUS and the 1A and 1 B Inverters?

1A DC BUS VOLTAGE reads approximately (1) 1A and 1 B INVERTER AMPERES are reading approximately (2) A. (1) 0 DC VOLTS. (2) 25 amps and being powered from the bypass source. B. (1) 0 DC VOLTS. (2) 0 amps and being powered from the normal source. C. (1) 125 DC VOLTS. (2) 0 amps and being powered from the bypass source. (1) 125 DC VOLTS. (2) 25 amps and being powered from the normal source. Page: 116 of200 12/1412009 explanation When the Battery output breaker is opened, LA-05, WC3 will come into alarm due to the b contact from breaker LA05. WC2 shows either a low voltage condition or a ground. In this case it would be a ground. The battery output breaker has opened due to a ground on the battery and when it opens WC2 clears. The annunciators provide indication that the breaker opened and the white light provides indication of the ground. For this set of circumstances, the battery is no longer aligned to the bus and the battery charger is carrying the load. The indications will remain normal and the inverters will have normal indications. The inverters will not swap to the bypass source and will still be powered from the BC. A -Incorrect. 0 DC volts on the 1A DC bus indicates the bus is de-energized. The bus still has power from the Batt. chger. The inverters will be powered from the BC or the normal supply and will indicate 25 amps. If it were to swap to the bypass source, it would still have amp readings, but if the manual bypass switch were to be placed in the bypass position, then the amps would be 0 amps. B -Incorrect. 0 is not correct for both. Normal is correct. C -Incorrect. 125 is correct. 0 is not correct and bypass is not correct. D -Correct. 125 is correct and 25 is correct from the normal source. DWNG: D177082 sheet 1 Page: 117 of 200 12/1412009 Previous NRC exam history if any: 2007 FNP NRC exam, this question is the only one in the bank tied to this KIA 063A3.01 063 D.C. Electrical Distribution A3 Ability to monitor automatic operation of the DC electrical system, including: (CFR: 41.7/45.5) A3.01 Meters, annunciators, dials, recorders, and indicating lights ............. 2.73.1 Match justification: It meets the KA in that it tests the ability to determine the proper readings on the EPB for an abnormal condition based on the indications and alarms received (white light and annunciators). The automatic portion of the KA is the breaker opening on an overcurrent condition. Objective:

6. DEFINE AND EVALUATE the operational implications of normal / abnormal plant or equipment conditions associated with the safe operation of the DC Distribution System components and equipment, to include the following (OPS-40204E07):

Normal control methods
Abnormal and Emergency Control Methods
Automatic actuation including setpoint
Protective isolations
Protective interlocks
Actions needed to mitigate the consequence of the abnormality Page: 118 of 200 12/14/2009 Date: 10r\Q009 I --1--------------------------

Time: "\02: 18 PM A B c D E r 12.0V AC PNl 1J BKR. #6 D-I77024 FU NEe (NOTE 12) FU ClF (NOTE 12) FU elF (NOTE 12) FU NEC 2 3 2-lie !II? TO 600V LOAD CENTER 10 1-5/c -I/O * ';Q* 11) ------+ I 1" 1'1. 5 v I I II I I I I I ;--(NOTE 7) I I --.--------" I "0 / " II :> " '" li! I I I I I I I I L I L

0;-----101 ':lHUt-n W CoOOA,IOOM 1 r'" (NOTE 12) FU NEC , (NOTE 12) -(NOTE. ') 07

§!l)l'l._ LAOe, §k :> u Q. N I TO 125 V DC 8US2.A AUTOMATIC TR,\NSFER SWITCH (NOTE 10) DIE5E.L GENE.RATOR Ie CONTRa L PANE.L *'* Gil '-4-1 LO(,,"-" 12.5 V DC DISTRIBUTION F'ANf.L IA 400 AMP MAII0 I U -2. '¢ , u N gt 0) "3 U g, .g N 4: (( uJ I:! uJ :> -z 5 6 I;>" p+-i II Y Cj} ,';... ,J. VI V A Iz:::, v..c::::J

n ':>EE I-lOTE i; gf Y ¥""'" '" CJl ill l-w > << '> 'i n ill It :J ::> IL ::> V .0 '" lL r:s: oJ,! ::< w '> 7. % '" U' r" / 9 IR 4ILOOI&-A 1-VC-.j"O WI'2S\l.Q.

own!.. eovJE'I'-"\1l t;WITCHI L_I----li G\:'4I1SQOII TO e:.j".! E.P.I::>. g'1; i j) 72 LA!? { <J TO IZ5V DC BLlS 2A AUTOMATIC TRANSFER SWITCH (NOTE 10) DI!,:5f_L -GE.NE.RATOA 1-'2.A CONT120l PAI-\EL -. .. I 'I.e; v DC DI5TRIe,UTION Pt>-NEL. Ie, 400 AMP MI>.IN Title: C:\Reference Disk\Exam Reference Disk\Drawin-gs\D177082-0001.cal Date: 1 8 <3 10 11 TO GoOOV LOAD TO GoOD V LOAD CEWrER I D CE.N.ER 1 E. -L SEE NOTE:3 j ,-3/C-I/0 -------, 1-81,,-1/0 S-I/C -I/O ..-J-.... D MCB J( '2-V'::::-*IO 101<20-1.QG BATTE.RY CHARGE.R 1 C f.OWSI'-11lAN.S. ,---+_--" C,"1I.0011>.-1>. TO e::,r* e>. E-P.t:>. I L ___ i;:========-- 2.--500 HeM CU Mc.!l 4-1/" -500 MCM c.u 2.-Z/C.-500M'CMGU'T' r TO B-TRAIN DISCONNECT I 0 I REF. DViG. D-I77083 "l 01 CoOO A 100 My L (RE.F, owe..

(';;EE NOTE. 3) 2{ NO ! ______ .J 811'Z.-L""1)

,?C BUS It>.. laO,? AMP \9 f _ .. J J }n -LAI'" 'I) 72* LAi7 )n-LA'2.0 aI I 0 oY ::> r-7 g E Ii ], { I,,,,V DC. Pt..NEL 1c..1 QIR4ILOOIC-A '::.'r 1*'2/C-4/0-'" 400Alv1PMA'" DI B.-A JT'ON Pt.-NE-L. "".1><,,," <J TO 1251/ DC BUS 2A AUTOMATIC TRANSFER SWITCH (NOTE 10) DIE.Sf_L I -'Z.I>. CONT120L PANoE.L"'" Ie. rt ISA. ISA WPP-I:>ORON REC'iC.LE. P"lL 'I '-"--AJV-QIP 5[001 A-A ¥.IPP LIQUID PIoiL _ rv FLOODINe. SENSORS -. 'I v-..--.,AFV . -MAIN FWP TRAIH A '20.0. H'iDR06EN REeD.,!\. 1-2/C-#60. 6 ,.." 6rv /I'2-I,t .c;, CONTROL P>lL CONI. PNL I A ---AJV . QI PI SI-lF';;SZ.c"O'A-A SOA .:lOA, SPAIZ.E. SPARE 30A '!lOA 5PARE SPI'-RE fiRE PROT. Fin-UNIT i-21c-""t'. '" """II /12-Yc#f" PROT. FILT. UtllTS CTMT, -Arv CONT. RM. FICTER UNITS SPACE I 14 SPACE SPACE IS 1(0 SPACE SPACE 17 18 SPACE SPACE .... SPACE SPACE '2.1 L1. SPACE SPACE '23 '21; SPACE Time 12 13 NOTES: If,lDIC.AT<=.O OTHERWI<;;E, AL.L. CIClCU'T BRE.AKERS 0\01 oc e.u,,:> IA ARE t' L,V. AIR CIRCUIT e,RE.A.KE..RS, "-POLE., 1.<;',000 AMP "l:C 2'5-0 VO L"'" D.C. P-Rt>.NE. SILE./TRIP RA11NG ARE-A';) SHOWN, '\01:38 PM A B c 2. ALL.. CIRCUIT BRE.AKER5 ON DISTRIBUTION PANELS EXCEPT THOSE NOTED ARE I D MOLDED CASE, TWO POLE, 100 AMP FRAME 10,000 ANIP Ie e 1'1-'5 VOC 3. T*HE.SE BREAI<.ERS IN D.C. DISTRI!!>UTION IS '::>0 THAT Ol-lL'i o .... e. '::>e.T OF BRE.AKEI<'5 C ... N 5E CLO<:.Et> AT ... l'IME. (5E.E owe;. 4.""'-OCN.OTES EQUIPMENT LOC.ATED IN DIE'l:E.L !:>LDc:., 5. DELE.TED '" "'-DE.\o.IOTE"" E:MER()ENC'I PDVJER BOARD IN M", N CONtROL. "'OOM. 7. 2 CONDUC.TOR<:. U'5EO FOR IE REMOIE WHITE-L.IG.kT AND 'C.Ol-lOUc..TOIl."" FOR REMOTE. VOLTME.TER. 8, REVISED 1'I>.IP COIL RA.TINCI ,.0 A.C;RE.E. WITH FIELD CH"",C;E TO BE MI>.DE., 9. Ie, I':> \-lOt,j -Q 10. SWITCH NON-S>E.LECTIVEL'I T-RAIJSFER BETWEEt-I THE TWO 12.51/. DC WHEN EITHER SOURCE FALLS BEL..OW 90% OF NOMINAL VOLTAGE _ (2-PDLE 125V DC 150 AMPS) 11. 1-4/C -#16 12. REFER TO FUSE MANUAL A-181987 FDR FUSE SIZE AND TYPE. F Title: C:\Reference Disk\Exam Reference Disk\Drawings\D177082-0001.cal Date: 10r -I I '2. ... 02 I E <>-(NOTE. ') 1\ ! 1 AUTOMATIC 87 TRANSFER SWITCH (NOTE 10) J u -S?. DIE5EL GENERATOR IC PANEL ",* .q. CONTROL , F G H J (NOTE. 11)-+

U N-, G! I LOOI A-A 125 V DC DISTRIBUTION PAN._L IA 400 AMP MAl\..} ISA ENlERG.lIGHlING
15A E.II\E.RG, LICo\o\TINC,

"",5". H'£OEI'. --"'-.A--ISA 51'1'1\\,£'" ISA REU,'1 RA.C.K,... IS'" TERMINATION CAB 4 _ (50l.. --., "'J\.A.. -11 lSA TERNI'N .... iION CAeS 2-1/"-(" ,':>01.. vd IS'" lER ..... INI>.TION ('SOL. W) ISA TERMIN ... TION CAe. "'. (SOL. 'Iv) IS'"' TERMIN .... 110" CA\)(P _ -v--Jl"_ (';)01.. W) -'I'-A--1-" IS'" BOP LNST. PNL IL-IM IS;A <-I/e. "(i,,, _ C.Ae.. I/>., "\I \..A./ ISA AUx. SAFf"U ... i1-0 _ -yo.,23" QIHIINC.ASC £$0(;.0*...,----- -15A

LIGHTING ISA FEEDER SPIIR£A A 2-11c.-"1O TERMI NATION CAe,., -.JJ"'V -('SOL V_) 20A I .. "1l.rv-_

'---"JQry _ TERMINATION G ... e. + -.IV V* -(SOL 'Iv) ISA CAe> & 113.0. Vv) /I2-1/C.-*" Te.RMINA.TION c.Aee (SOL V,) TE';R,MINATION 10 15A C'OOL y,) 2-1/'--"(" TE.RMINATION CAe. (., (':'01. V") ISA 15A TE.RMlt-IAlION CAe. c. (':>Ill 'i.) fl.ELA,( eM'> Ie, ISA /I '1.. "c.-t/O (AI? IE. ISA ""uf'v. CONTllDl.. Il£MOTf. <;r ... TlotoJ 2-IIC #10 I\. -v<§" NSHIINGLCR2507J-N -"'-"--ISA A..rV QIHt'l. LOOII:-A I'SA TERMINATION CABINET, TERM. CABINET 4A FOR HCB PNL 2 _ QIH25L004-A -'\['"\,.A -IS"' 50A ,,-CIRlUlT BRE?I'-ERS YHt,Jl 10 UtlUSED C.1'I"bL"S TI-II'I, IN T£I'.\'I\.1'.O'I. I'IITM53 1 2 3 4 i <{ Cl uI uJ l-I-J cr: Cl Ll uJ W ;;> > 0 ---v , 4. <! u > ---. > N "-Y. !.. lL c( uJ l:< '" > :z " '" Time \03:29 PM ! ! I AUTOMATIC TRANSFER. SWITCH (NOTE (0) DIE-.Sf_1.. -GENERA10R 1-'2,t>. CONTeOI. PA"'EL .... Ul '" r: ,..: !"Z.e; V DC G\ I R41 LC>DI e.-A DI':>TRIBUTION PANEL Ie. 400 .... MP Mp..IN SOA 4.I(,KV ':.WGI'. BlJ':> (INC.. BKR "CON,) SOA 4.1("K,V SWGD. 8U'S IF __ v ... r-... (I'OR BKRo;, CO"i) PENETaATiOt-I ?oM 30A l'SOLATlO" 2-1/C-#6",_ I V\.-50A 4.I"'KV 15A CRDM. M-e, SET 1.0. _ r--. GONT. CAe.. " v'-/C " 30" 4*'" K-I "'GR. e,u", IF _ V"\!L" TE.':.1 CAB. -",-,,--" ISA 4.1(.,KV 'OVJGI2 LOCAL c.ol-n. PNI... IS A 4.I("KV 5VJG2 e,U';IH LOCA l.. CO"T. P"L LOCAL HDTSHUT DOWN PNL. lA .. rJl" QIHZINSM*PV.OS .... A -V'V'-TERMINATION CAB. lEo.M. (,1>. QIH!:'ltlOrrA ! ., '20A 'OPALE. 2" 2S eOA r-...1:rx:..---v1 2-1/C #6\0001/ LOAO CENTER 30A I A (I NC.. I:>KI<. CONT ) -A 2,-lIC #6G>00V LOAD C.EKTER A..rV I'" \FDR CONT) "/V'" A 2-lIC.J!2.(D00V 1..0AO C.2.NTEIZ 30A-A.rV -tD (INC. e.KR. '-ON" \ 30A '-"'10,.. A 1-2/C #4 G>OOv LOAD GENTEI2 -=)'J'V -I R(INC. BI<R, C.ONT) ('\ 17rv-/I 2-lIC #6 HOTSHUTDOWN PHI... AUX RL'i --...II...rV -C"'SItJET 20A ""J1rv- ... .AJV ---SE.QUENGER BI F A 2-lIC 116 RIO.A.GfOR TRIP -.,A..rV -'Ov.x.;R. NO. I :lOA flirt 15A r-. ilCry _ 4.16KV SYlGR. BUS Ie . -.AJV ---WV & UF RELAYS) 20A 24 "",PAC.e. 2c. ':>P ..... c.c:. 'SPA-ee: !OPACE SPA.CE. -oPACE. 5 6 7 Title: C:\Reference Disk\Exam Reference Disk\Drawings\D177082-0001.cal Date: 1 1 AUTOMATlC 1 Time: r -'104:0B PM LUL.A. 1 t: LJ I N TRANSFER. SWITCH (NOTE 10) DI't:5f_L 1-2.A CONTeOl.. 1< ... QI&-A JTIQN Pb,NEL Ie. ""I><IN -x 'V"" 2-1/C 1I6(.oov LOAD CENIER -IA (INC, I>K.R, COWl) -v-_ /LZ::l.lC_1I6<;,oov LOAD CENTER A_rv BI<<.R"'. CONT) -v-.. A LOAI) .JI....TV -10 (INC BKR. CO"" \ -v-LOAD C!:NTE.e ..AJV -. ID (FD1a.BKRS, CONT) ,-/11-2/C 1I4c.,OQV LOAD CENIEI? )v-v -I I< (IN(.. ol<R. CONT) -v-/1 2-1/C 1I6 HOT 5HUTDOWN PHL. RL'i ..I\J""V -CABIIJET QINeILOOIA-A "Y __ .AJV .... .. ... ._-SEQUENCER e,1 F -v-/1 2-1/C 1I6 RUCTOR TRIP ..A/"V -':>v.JGR. NO. I elH '"'Y _ /1£:1&1112 4.16KV SViGR. BUS lC .AJV -(UV !. UF RELAYS) SPARE !>PACE. "Pt>-CE. "PACE ,PACE. .... ,...,. ..... ' ..... .-"' .... 'FWP A 20A. '20.&.. H'IDR06EN RECOIll!O, '"" 6rv /l2-1jc. "rp SAMPLE CO>lTROL PIolL c.eNT. PNL I A -"I'--A.---...A..FV ' QI P I S>lFSS2.G>OlA-A 30.... .:lOA SPAeE. SPARE: 30A '!>O .... SPARE 5PP-RE FIRE PROT. FILT. uNIT j-2!c-#-/s,,,, /l2-Yc#b FIRE PROT. FILT. UIIITS CTMT, -AJ'V" -CONT. RM.FILTER UNITS SPACE 14 SPACE SPACE IS ". SPACE SPACE 11 Ie SPACE SPACE '2.0 SPACE SPACE '1.1 1.1. SPACE SPACE 23 '24 SPACE SPACE 2S '210 SPACE SPACE 21 28 SPACE SPACE SPACE 1"0 STATION 5ATRRi 10 I'l.SV DC DI<;"RIBU\lON ?AlIE-L Ie; I'Z.SV D,('.(TLJRf.>INE.. f.>l..DG) 400 AMP MAIN. 1-2/C-1I2 o,E.E. '" 50".3OA _ 4.I<iKV 5'lX;12,£OU5 I .... LOAD CENTEI< (INC, !>KR.CQNT) SOA 50A II> (INC. £OK". CO"T) LOAD CENTE\: (,D\<, \:>KR. CONi)

50" 00'-If:> (FDR.. to'l-.R CONT) 4,I"KVSWGI'.,B\JSIC 2-1/"'*<<>

,.-....!!! ... c:;,OOV LOA\)CENTE12 !>KR. COIlT) 50A eeA 1M (INC, BK.I!., CO",) 4.16 KV S\lGR. BUS IC Z-I,t!",., 7 .-.. C-!2-v _ C.OOV LOAD CE,'nEe (FOR SKR. CeNT) ao.. 30'" AJ'V" -1M l INC, FD<<. P.>I<il-eOi'm £to ...... "7\ ut.f'(U I t:.I,JUIf-'MI:.N I eoLO",. 5,UELETED <<> >/0-E.MEp'C;ENC'/ PDWE.R BOARD It-! MA\N CO""ROL 12.00M. 7. 1-4/C -"12: 2 CONDUC TOR'=> USEO FOR RE MOTE WHITE. LIGHT AND '2 COt-lOUC TOR", USED FOR REMOTE VOLTME,TER..

e. REVISED il'\IP COIL RA.TIN<:>

,0 AGREE. WITI-I FIELD CH"'NGE. TO BE MADe. 'I. Ie. Ie; I-IOlo.l-Q

10. SWITCH NOIo!-SE.LECTIVELY T-RAIISFER BETWEEJoI THE. TWO 12.5V. DC SOURCES WHEN EITHER SOURCE FALLS BELOW 90% OF NOMINAL VOLTAGE. (2-PDLE 125V DC 150 AMPS) 11. 1-4/C -#16 12. REFER TO FUSE MANUAL A-181987 FOR FUSE SIZE AND TYPE. REFERENCE DRA 'WINGS: D-ll'1000 SINGI.!: LINE E.LECTRICAL

"'U'/.ILlAR'I (NORMAL 41(QOV G (.,oov) c-nlOOI D-I77024 SINGLE Ut<E. E LEC TRIC.AL ... uJ<: I LIAR.'1 (E.ME RGE NCoY 411.0V c"OOV) SINGLE LINE, 120V AC VITAL REGULATED S'I5TEM />.. A-177538 El..E.CTRICM_ c._NaP-AI. DETAIl..':. 4 NOTE'<> 0-111083 SINGLE LINE DC DISTRIBUTION S'ISTE.M Ie. C-I'17133 INTERLOCK '<>CHEMA.Tlc BATTER,( c:= =:: 41.0 U-175725 DC S\lGR. LA05 DC GROUND DETECTOR U-175727 DC SViGR. LA02 CONTROL CIRCUIT 4*1(01<'1 '=>VJGR Moo _ Y-\!./""', eUILOING I A TE"" CAB 'I '-A ;lOA .!Ioc.. FIRE PQ.OTE.CTIOIol, IJIY4'!>5010A-1l SMO\(¥. OETe.::.roil-1.-I/c..i.p I' V"'>l!."-'" ""' 11rv-/l'l.' 1/ .... r:,op PANE-I... P SYS. INVERTER "'-"----.J.....rV G.IHII>JC;i?'i.So4f'-N -N sop.. 50A SPARE '5PAeE. 2M 15A WASTE COOl-Ell VIP.>Il."IUITC.HES NIGi2.4N9WBP2(P14-N 15A 15A NITe.141 oERVICE 1\.. _ V"\!2.------ "-"'W.rv-/l'1..-I/('.'(" FIRE PROTECTION lGC!\l "TAIIOJ.\ 'I '-"---.A.rV' -FILTER UNIT t;\v41 LOO::\-\..\ 15A 15A FIRE PROTe.C.iIO'" _ """' 1'11>.1:. PIl.OTi;.c.Tlo'i

O'1'S. tJITe,oQi:> . "I'-A--"PIl.IIJJ'.LE.R SPACE: 21 Z2 ':>PACE @ 01770821 [yVY'2002 I OCP-01 I ""PAC E.. Southern Company Services, Inc. for "'RE PROTELT'O\oJ 2-'k,"h /I.

5PRIJ.lF'le.R 5;","'118214

v "-A. --

PRDTe;CTION 2-lk #" .... 26 SPR1"l<'Le;R

>)'S. I>lrrB219-V

-........ -SPAC.e. 1::,ode"""',;, ALABAMA POVER COMPANY E I--F I--G I--H I--10< us. onIj by employees

m.

m The J M FARLEY NUCLEAR P' 'NT _ UNIT NO 1 Southern Company. Unautho/'lzed po$$CSSfOfI. use. lidtribution, copymg. I'ttDXCT'

LJ'\
J disoomirootion.

0< <he"""" m Ofly portion hmof ;, prohibiled. SINGLE LINE ""I SIJ'V I .1 I ** 1_1 .. ' .. 1 ..... 1_1 I. .1_1"",,1."1'1 I. .1 ...... 1 ...... I"""I""vl. I. .1"",1 .. ""I .. ""I, .. vl. MA BT SU>..<<:T tuTltNGltI1tCH11Y ,CIt('lIltNGItlllEPTIDEP1'IDGtIKCt1 IV ,ta:'DIDGtllll:Ptl>>O"TIOGtII£CH )Y ,tHK'DIEIGtIEVTllIEPTIDOtIJttH ", ICtIC'llIDColtIIlEPTlreTJ[trQllttH DRAVN __ . _'_ Met[]) _._._ 1a'0<0> ___ D.C. DISTRIBUTION SYSTEM lA 1 1 -I-I I I I I I I I I I I I I I I I I I OCP I MLH I DEW I "'1:::::>1<::1" H'I'R!N£D M. MALCOM "'Tt 12/4170 t NONE "" VER. N:l __ J)4T[ VtR. NO. -->>AT( VEt. NO. --DllTE VElt. NO. ....+/-!:Q.... 1Iro\1t 06/07 /06 MTE SHEtT 1 [F -Si£Ea 1 REVISED PER ..sN ---10<4-9005501E002, VER 1.0 N'1'MIVD Oft,TE ___ ..... stIltS D-177082 I 7 8 I 9 I 10 I 11 I 12 I 13 OPERATIONS CRITICAL DRAWING Title: C:\Reference Disk\Exam Reference Disk\Drawings\D1770B2-0001.cal

45. 064A4.03 OOlINEW/RO/MEM 3.2/3.3/N/N/3/CVRlVER 5 EDITORIAL The 1 B DG is being paralleled with the grid for surveillance testing, and prior to closing the 1 B DG output breaker, the Synchroscope is turning fast in the FAST direction.

Which one of the following states: 1) the component with the highest frequency (the 1 B DG output or 1 G 4160V Bus), and 2) the direction that the GOVERNOR MOTOR SPEED/MW switch must be turned to adjust frequency prior to closing the output breaker? (1 ) (2) A. 1 G Bus Frequency RAISE B. 1 G Bus Frequency LOWER C. 1 B DG Frequency RAISE 1 B DG Frequency LOWER A -Incorrect. Both parts are incorrect, and are the exact opposite of the correct. Plausible, since confusion could exist as to the relationship in this scenario between the two frequencies. Also, this choice would be correct for this indication if the DG was on the bus and the Grid was being paralleled on after an LOSP. B -Incorrect. The first part is incorrect (see D). Plausible, since it would be correct if the synchroscope was traveling in the other direction for this condition, or if the DG was powering the bus and off site power was being restored to the bus. The LOWER is correct for the given conditions. C -Incorrect. The first part is correct, but the second part is incorrect. Plausible, since it would be correct if the synchroscope was either going too slow in the same direction, or traveling at any speed in the other direction. D -Correct. The oncoming generator must be at a higher frequency output to turn the Synchroscope in the clockwise (fast) direction. It must be going slow in the "fast" direction prior to closing the breaker per STP-80.1 Step 5.9. In this case, the DG must be slowed down by going to LOWER on the Speed switch until the synchroscope is traveling slower in the same direction. STP-80.1 Version 47 Page: 119 of 200 12/1412009 Previous NRC exam history if any: 064A4.03 064 Emergency Diesel Generators A4 Ability to manually operate and/or monitor in the control room: (CFR: 41.7 /45.5 to 45.8) A4.03 Synchroscope ................................................. 3.2 3.3 Match justification: To correctly answer this question, knowledge is required of monitoring the synchroscope operation, and interpreting what information the synchroscope is communicating. Interpreting this information from the syncroscope properly is required to determine how to adjust DG speed for paralleling the output breaker safely. Objective:

6. DEFINE AND EVALUATE the operational implications of normal / abnormal plant or equipment conditions associated with the safe operation of the Diesel Generator and Auxiliaries System components and equipment, to include the following (OPS-40 1 02C07):

Normal control methods
Abnormal and Emergency Control Methods
Automatic actuation including setpoint (example SI, Phase A, Phase B, MSLlAS, LOSP, SG level)
Protective isolations such as high flow, low pressure, low level including set point
Protective interlocks
Actions needed to mitigate the consequence of the abnormality Page: 120 of200 1211412009 05112/09 12:44:19 FNP-I-STP-80.1 NOTE: To prevent entering an LCO, Step 5.8 should be completed in its entirety.

If the entire step can NOT be completed, the Shift Supervisor shall declare the DIG inoperable at the time of synchronization (Step 5.9) and enter the applicable LCO. (CR 2008105195) 5.8 To ensure the DO load is within the 300 hour rating (4474kW) in the event an LOSP occurs with the DO connected to the bus, verify the following equipment not in service: [ ] B SW Screen Wash Pump IF 1-2L LC aligned to Unit 1 [] B RW Screen Wash Pump IF 1-2J LC aligned to Unit 1 [ ] MDFP [ ] #4 RWPump [ ] #5 RW Pump NOTE: When running two DGs simultaneously, DGs should NOT be synchronized and tied to the same unit. For routine surveillance the diesel is normally aligned to the unit number corresponding to the air header number aligned as indicated in the surveillance schedule.

5.9 Synchronize IB Diesel as follows: 5.9.1 Verify an operator is standing by at D008-1 for breaker observation.

5.9.2 Place SYNCH SWITCH for IB DO output breaker (D008-1) in MAN position. NOTE: Initialing for completion of Steps 5.9.3 and 5.9.4 may be done after Step 5.9.5. 5.9.3 Establish and maintain the following conditions until Step 5.9.4 is completed:

Adjust generator voltage to match running voltage.
Adjust generator frequency to establish a slow synchroscope speed in the FAST direction.

to ... .lUhe switch for the output breakerand.the

- - goy;ernormotp.rs-witch be >. given to utilizing . two operators fortheJollm"ingtw()steps
.(OR 1-2000-282) 5.9.4 Just prior to the synchroscope reaching 12:00 position, close IB DO output breaker. Version 46.0

46. 065AK3.04 OOIIFNP BANKIROICIA 3.0/3.2/N/N/2/CVRIY A complete loss of instrument air has occurred on Unit 1, and the following conditions exist:

The Reactor was tripped from 100% power.
The TDAFW pump auto started.
BOTH MDAFW pumps failed to start.
SG NR Levels are slowly trending up and read: 1A: 27%, 1 B: 29%, 1 C: 30%
Instrument Air is expected to be lost for the next 4 hours while repairs are made. Which one of the following describes the action(s) required for the TDAFW system that must be taken and the reason? A. Use the jacking device to open HV-3228A, B, and C, TDAFWP TO 1A, 1 B, AND 1 C SG FLOW CONT valves, to ensure an adequate heat sink. B:I Start the emergency air compressors and align the EACs to supply air to HV-3235A and HV-3235B, TDAFWP STM SUPP valves, to ensure an adequate heat sink. C. Use the jacking device to close HV-3235A and HV-3235B, TDAFWP STM SUPP valves, in the main steam valve room (MSVR) to prevent an uncontrolled cooldown.

D. Start the emergency air compressors and align the EACs to supply air to HV-3228A, B, and C, TDAFWP TO 1A, 1 B, AND 1C SG FLOW CONT valves, to prevent an uncontrolled cooldown. A -Incorrect. The AFW FCV's fail open, and do not need to be jacked open to provide an adequate heat sink. Plausible, since they do need to be jacked closed or throttled when necessary during a loss of air. Also, confusion may exist between the steam admission valves which do fail closed (after 2 hours) and the FCVs which fail open. B -Correct. Per AOP-6.0, Version 35, Step 8 below. For the first two hours after the loss of air, the installed accumulators maintain the steam admission vavles open, but after that the emergency air compressors must be started to supply them with air to maintain them open, maintain TDAFW pump speed, and to maintain AFW flow to the SGs for a heat sink. AOP-S.O, Version 35, Step 8 8 Maintain SG narrow range levels between 35-69%. *************************************************************************** CAUTION: The TDAFW Pump steam admission valves will fail closed within two hours if emergency air is not aligned. ************************************************************************** 8.1 AlER WHEN TDAFW Pump is started, THEN vary TDAFW Pump Speed to control AFW flow. Page: 121 of200 TDAFWP SPEED CONT [ ] SIC 3405 adjusted 12/14/2009 8.1 RNO IF the TDAFW Pump steam admission valves fail closed, THEN align emergency air using FNP-I-SOP-62.0, EMERGENCY AIR SYSTEM. C-Incorrect. With only one AFW pump available and SG levels below 31 %, at least 395 gpm TDAFW flow is required. Closing the steam admission valves would not be allowed in this case. This is plausible, since the SG levels are trending up and close to the levels at which AFW flow can be secured. Also, jacking devices can be used to close the valves such as isolating a faulted or ruptured SG, and at Beginning of Core Life cooldown would be excessive with full TDAFW Flow. Also, throttling with the jacking devices for the FCVs would be an option to limit excessive cooldown if necessary. However, AOP-6.0 would direct reducing TDAFW pump speed from the MCB Pot to control the amount of AFW Flow to the SGs. With no Instrument air for 4 hours, decay heat would require steaming the SGs and makeup from the only source of AFW to the SGs would continue to be required even after SG levels were> 31 % NR. D -Incorrect. The emergency air compressors must be started by for the Steam Admission valves which have air accumulators keeping them open for at least 2 hours, but cannot be assumed to last for 4 hours. Plausible, since confusion may exist as to which of the the TDAFW valves are supplied by the emergency air compressors and which valves fail open and which valves have 2 hour rated air accumulators to keep them open with a loss of air. The UPS does supply the HV3228 valve solenoids for up to two hours, but the emergency air does not supply HV3228s. AFW FSD, A181010 3.14.7.1 The emergency air system shall provide a backup air source for these valves as a means of additional reliability in admitting steam to the turbine for TDAFW pump operation (References 6.7.049, 6.7.050). 3.14.7.2 The instrument air system shall supply clean, dry air at a range of 80 to 100 psig to the air reservoir for steam supply isolation valve operation (Reference 6.5.001). 3.22 TDAFW PUMP UPS SYSTEM TPNS No. QN23LOOI-AB (UPS) and QN23EOOI-AB (Battery) 3.22.1 Basic Function The UPS system shall be designed to provide an un interruptible source of 120 V ac and 125 V dc power supply for control of the TDAFW pump turbine drive (QN23P003), steam admission valve (QNI2HV3226), steam supply isolation valves (QNI2HV3235A, B), instrument air valves (NN12SV3412A, B) and the TDAFW pump discharge flow control valves (QN23HV3228A, B, C) for a minimum period of 2 hours considering loss of both offsite and the backup diesel power to the UPS (Reference 6.7.074). D175035L (Emer air compressors) D175033 sh. 2 (D-9) Air supply to SG Atmosph & TDAFWP Stm Admission valves Page: 122 of 200 12/1412009 Previous NRC exam history if any: 065AK3.04 065 Loss of Instrument Air AK3. Knowledge of the reasons for the following responses as they apply to the Loss ofInstrument Air: (CFR 41.5,41.10/45.6/45.13) AK3.04 Cross-over to backup air supplies ................................... 3.0 3.2 Match justification: AOP-6, Knowledge of the reasons for Cross-over to backup air supplies [1C or 2C AC, N2 to PORVs, Emerg. AC's for SG ATMOSPERICS] as they apply to Loss of Instrument Air. This question asks for the action and reason for the action on loss of instrument air. The reason for starting the Emergency air compressors in the given scenario is that SG makeup is needed for a heat sink and the other options will not provide an adequate heat sink for 4 hours. Objective: AOP-6.0 1 STATE AND EXPLAIN the operational implications for all Cautions, Notes, and Actions associated with AOP-6.0, Loss ofInstrument Air. (OPS-52520F03)

2. EVALUATE plant conditions and DETERMINE if any system components need to be operated while performing AOP-6.0, Loss ofInstrument Air. (OPS-52520F06)

Page: 123 of 200 12/14/2009 07/02/096:29:26 FNP-1-AOP-6.0 LOSS OF INSTRUMENT AIR Version 34.0 Action/Expected Response 7 Maintain PRZR level between 20-50%. 7.1 Alternately cycle open and closed one of the following MOVs for charging control as required. CHG PUMPS TO REGENERATIVE HX [] QIE21MOVS107 [] QIE21MOVSlOS 8 Maintain SG narrow range levels between 35-69%. Response Not Obtained 7.1 Go to FNP-1-AOP-16.0, CVCS MALFUNCTION while continuing with this procedure.

CAUTION: The TDAFW Pump steam admission valves will fail closed within two hours if emergency air is not aligned. ------*************************

- - ***********************************************************

S.I WHEN TDAFW Pump is started, THEN vary TDAFW Pump Speed to control AFW flow. TDAFWP SPEEDCONT [] SIC 3405 adjusted Page 4 of 12 S.l -------- 10122/2009 14:29 FNP-I-ESP-O.l 1 It It REACTOR TRIP RESPONSE Step Action/Expected Response n Revision 29 Response NOT Obtained ************************************************************************************** CAUTION: To provide adequate heat sink total AFW flow must remain greater than 395 gpm until at least one SG narrow range level is greater than 31%. _Page Completed SG MDAFWP TO lA(lB,IC) QIN23HV MDAFWP TO lA(lB,IC) FLOW CONT HIC AFW FLOW TO lA(lB,IC) SG [] FI 3229A [] FI 3229B [] FI 3229C AFW TOTAL FLOW [] FI 3229

Control MDAFWP flow. MDAFWP FCV 3227 RESET [] A TRN reset [] B TRN reset MDAFWP TO lA/lB/lC SG B TRN [] FCV 3227 in MOD lA IB SG []3227 A [] 3227B in MOD in MOD SG []3227AA [] 3227BA adjusted adjusted lC []3227C in MOD []3227CA adjusted Step 1 continued on next page. Page 5 of 37 07/02/096:29:26 FNP-1-AOP-6.0 LOSS OF INSTRUMENT AIR Version 34.0 TABLE 1 COMPONENT MANUAL OPERATOR NUMBER NAME OPERATOR DRAWING NIN21V870 SGFP1AINBOARDSEALPRESSREG NO OPEN red NIN21V901 SJAEBYPFCV YES OPEN A ; hio N1N21V902 GS COND BYPASS FCV YES AS IS NIN21V908 CNDS MINIMUM FLOW FCV YES OPEN NIN21V909A SFGP lA RECIRC FCV YES OPEN U-213892 NIN21V909B SFGP IB RECIRC FCV YES OPEN U-161476 NIN21V916 CONDENSATE PUMPS BACK UP YES OPEN COOLING WATER PCV N1N22V725 SGFP SEAL DRAIN TANK AUTO DUMP YES CLOSED TO THE CONDENSER QIN23FCV3227A MDAFW PUMP TO STM GEN 1A YES OPEN U-176885 (l-AFW-FCV-3227A)

QIN23FCV3227B MDAFW PUMP TO STM GEN lB YES OPEN U-176885 (l-AFW-FCV-3227B) Q1N23FCV3227C MDAFW PUMP TO STM GEN 1 C YES OPEN U-176885 (l-AFW-FCV-3227C) QIN23FCV3228A TDAFW PUMP TO SG 1A YES U-176884 (l-AFW-FCV-3228A) Q1N23FCV3228B TDAFW PUMP TO SG IB YES U-176884 (l-AFW-FCV-3228B) QIN23FCV3228C TDAFW PUMP TO SG 1 C YES U-176884 (l-AFW-FCV-3228C) Q1N25V001A CHEM ADD TO 1A SG ISO YES CLOSED (l-CI-HV-3772A) Q1N25VOOlB CHEM ADD TO lB SG ISO YES CLOSED (l-CI-HV-3772B) Q1N25V001C CHEM ADD TO 1 C SG ISO YES CLOSED (l-CI-HV-3772C) NIN26V887A HTR DRN PUMP 1 A RECIRC YES OPEN Page 17000 Date: 101': '7-009 Q1P18COO2A-A EMERGENCY AIR COMPRESSOR fOR MAIN STEAM ATMOSPHERIC RELIEF VALVES tPSV\ (&) P Q1PI8CQ02B-B EMERGENCY AIR COMPRESSOR roR MAIN STEAM ATMOSPHER!C RElIEF '/ALvES 1" X 3/4" RED. NV071.:" NV07St., c:t-;1'1-n_--<0-175034 SH.l(' 2)( I , o m or L ___ /_. NV072 I NOTE 4 3/4" X 1/2" RED. X 3/4" RED. NV0718 NvQ75B <. r3/4" x 1/2" RED. NVQ73A NOTE 8 r 3/4" x 1/2" RED ..___1/2" H8D-643 NOTE 2 -175034 (D-2 PI \NOTE 3 & 8 .2872AB f D-175033 SH.2(HiI> 70 __ --'j 1/2" DR. NV097A X RED. I 'NV0738 NOTE 8 ,3/4" X 1/2" RED. tP C0 , \E:.88SB 1/2" HBD-543 NOTE 2 -rp;9-FW /raW\ ,vJ<J -175034 (0-2 I D-175033 IV NV0958 L--_ .. ___ " 1/2" DR NV0978 ,-pIt FW , vJ\} Q1Nl1v016A SV .3371A8 )0-175034 r3/4" X 1/2" RED. NV073C @c NOTE 8 NOTE 2 t t-t '"' . Qt Nv076A NV076C Time 3/8" S.S. TUS. ,_/ 11 rPV\ s j-J ---I 3371A I Vl NOTE 8 ;:i Fe _ @5TI3SH fu '" 1/2" X 3/8" RED. 3/8" S.S. TUB. py ! ____ r-NOTE 8 ... , ___ Fe r-- d: 1/2" X 3/B" RED. r ;::-w s \47:24 PM 1/2" X 3/8" RED. 'f F ____ -NOTE 9 '---1/2" X 3/8" RED. NOTE 7 r7 3/8" S5. TUB. 7 @CA IS [>I()..-., ,fPYL

-NOTE 8 B'i ----I.e NOTE 9 "-,/2" X 3/8" RED. N ;::-____ -NOTE 9 1/2" X 3/8" RED. I Title: C:\Reference Disk\Exam Reference Disk\Drawings\D175035-0002.cal Date: 10r 'fOOg EOUENCER)

1 I ,TA )N /\ /'C7 , I y 323/4B e I A/S Y2214D I --1 v I I L 3234B Time: "")52:31 PM /2h 1" EBC-1 1" EBB-2 OV0048 J;@B HS-L--i---TDAFP START 3304BC +

HS___ I -, @BB I rPSV\ NIPI9VI488 3" RED

I '"

SV,\ I )8 I I 1-I I I I LO I 3235B I -.l S I OVO L 4" x L ---l HV 8-1 35B o I 1 HS_L / 'c ___ , : --TDAFP START '8-/1 1 HS:J C/ , ,/ : 3304AB-I -oUI 1 L -----1 N1P19V148A / 'L -i

1 3235At",.

VENT , / 1 1 : 3235A 1 " ------------- --i <I I RSVR. I (s\7"\ -l " A/S :J .. :J N1P19V1518 \ N1 P19V1478 Y L& I I << "'<t PSL v\-o-.f G 3/4" V OV0148 3" E88-2 OV014A OV0058 PSL\ :; 3412AA 0.-\ :J Z 3412A j ) ". ii' 1 A/ S Title: C:\Reference Disk\Exam Reference Disk\Drawings\D175033-0002.cal FNP Units 1 & 2 3.21.7.2 3.21.7.3 3.21.7.4 AFWSYSTEMS makeup water from the demineralized water storage tank (References 6.4.049, 6.4.051, 6.4.063). A-18101O The CST shall interface with the condensate system so that condensate water may be used as a makeup water source (References 6.4.049, 6.4.051). The recirculation line from the AFWS shall connect to the tank at 19 feet above the base of the tank. This location is above the portion of the tank dedicated to the AFW emergency supply (References 6.4.050, 6.5.008). The freeze line protection for the CST piping and instrumentation lines shall be provided with nonsafety-related power from the 120-208 V ac, distribution cabinets 1 CC and 2CC (References 6.4.064, 6.4.065). 3.22 TDAFW PUMP UPS SYSTEM TPNS No. QN23LOOI-AB (UPS) and QN23E001-AB (Battery) 3.22.1 Basic Function The UPS system shall be designed to provide an uninterruptible source of 120 V ac and 125 V dc power supply for control of the TDAFW pump turbine drive (QN23P003), steam admission valve (QN12HV3226), steam supply isolation valves (QN12HV3235A, B), instrument air valves (NN12SV3412A, B) and the TDAFW pump discharge flow control valves (QN23HV3228A, B, C) for a minimum period of 2 hours considering loss of both offsite and the backup diesel power to the UPS (Reference 6.7.074). The UPS system is provided with additional component redundancy to enhance the reliability of the UPS system. This redundancy includes an alternate (backup) UPS system (Battery Charger, Inverter, and Rectifier) located inside the same cabinet (Reference 6.5.015). 3.22.2 Functional Requirements 3.22.2.1 3.22.2.2 The UPS shall normally be supplied from an emergency diesel generator-backed 208 V ac single phase 60 Hz source (References 6.4.002, 6.4.045). In the event of inverter failure, the supply to all components shall automatically transfer within 50 msecs to a step-down transformer (within UPS) powered from the ac source (Reference 6.7.075). 3-49 Ver. 15.0 Date: 1 or; 6 I Time: ('\59: 11 PM I 7 8 j 9 10 . __ 11 __ ----.-1/ 12 \ AB 3A h >,< 1---, ?-\: VI / "I I I I I I V001A '<1"1 LC I V009A I 1 °1 ".".1 61 " L-=_ I '<1"1 )iA T I <..i . 1 Cf)UI i I r-L---L-----3D II ".".1 I L----l NV004 4" HBD-47 I ... , NV007 '<I" I 5B 'I ,/ <...> 0 m <..i 0 Et =...r (/) L<JV009B :-4 V060 > , <D '<I" '<I" '-...0 "'> 4" DBC-2 V045A V059A 'b'b'QO /H )'-C' .-:.Y, / ,/y, -c __ , ___ -:J-[BJ : MAlN FEED -...1 I I WATER 5--..., ...... -----------, DBC*I,DBB @ : I I I 4" DBB-2 I I ," DBB-l I I STEAM GEN t__ NO.1A ci W 0:: 00 x b YO'" I : )J 0 tM2;j' l"DR 1 -.: Q(f f /" -V047C 1" DR. V054C M 3350B . /7>o)f DBB I I I I I I STEAM GEN NO.1B NOTES: 1. THE SYSTf NUCLEAR THIS DRAV THE IDENl . COMPONEt ARE SHOll THE VALVE 2. FOR P&I 1 & 2. 3.

MINIMl PUMP MF(

47. 068K6.10 00 I/MOD/ROICI A 2.S12.9/N/N/3/CVRlY A Unit 1 #2 Waste Monitor Tank release to the environment is in progress lAW a Liquid Waste release permit and SOP-50.1 , Appendix 2, Waste Monitor Tank #2 Release to the Environment.

FH2, RMS CH FAILURE, comes into alarm.
R-18, UQ WASTE DISCH, is indicating normal on the Radiation Monitoring system console and on the recorder for R-18, RR0200.
The HIGH Alarm and LOW Alarm red lights are illuminated.
The control power fuse is found to be illuminated on R-18. 1) What effect, if any, does this condition have on RCV-18, and 2) what action(s) is(are) required lAW SOP-50.1, Liquid Waste Processing System Liquid Waste Release From Waste Monitor Tank? A. 1) RCV-18 will NOT automatically close. 2) Close RCV-18 with the Handswitch on the Liquid Waste Panel. By 1) RCV-18 automatically closes. 2) Verify RCV-18 closed at the Handswitch on the Liquid Waste Panel. C. 1) RCV-18 will NOT automatically close. 2) Close the manual discharge valve to the environment.

D. 1) RCV-18 automatically closes. 2) Override RCV-18 in the open position with the manual jacking device, implement ODCM actions for an inoperable R-18, and continue the release. Page: 124 of 200 12/14/2009 A -Incorrect. First part is incorrect, since the control power has been lost and will initiate the automatic action the same as if a valid high rad alarm condition existed. Plausible, since no high alarm condition exists, and the meter is reading a normal reading. If this condition was present, the second part would be correct due to the inoperability of R-18 per SOP-50.1 Step 3.2: "IF R-18 becomes inoperable while discharging liquid waste to the river, THEN the discharge shall be stopped immediately and the Shift Support Supervisor notified." B -Correct. Control power being lost causes a fail safe automatic function of closing RCV-18. The second part is correct per SOP-50.1 Step 3.2: "IF R-18 becomes inoperable while discharging liquid waste to the river, THEN the discharge shall be stopped immediately and the Shift Support Supervisor notified." C -Incorrect. The first part is incorrect (see A). The second part is incorrect since the release must be stopped immediately per SOP-50.1 Step 3.2: "IF R-18 becomes inoperable while discharging liquid waste to the river, THEN the discharge shall be stopped immediately and the Shift Support Supervisor notified." Plausible, since after ODCM actions have been implemented, the release could be continued with R-18 inoperable, but not prior to them being implemented. D -Incorrect. The first part is correct (see B). The second part is incorrect, since there is no manual operator on RCV-18. Plausible, since many AOVs have manual operators, and after the ODCM actions have been taken, the release may be continued with an inoperable R-18. The valve would be overridden open, but most likely the handswitch would be taken to open with jumpers installed to defeat the close signal. FNP-l-SOP-50.1, Version 66.0 3.2 IF R-18 becomes inoperable while discharging liquid waste to the river, THEN the discharge shall be stopped immediately and the Shift Support Supervisor notified. FNP-I-ARP-1.6, RMS CH FAILURE, FH2, Version 64.0 1. The radiation monitors fail to a "High Radiation" condition on loss of instrument and/or control power that will result in actuation of associated automatic functions. Refer to annunciator FHl for automatic actions. FNP-I-ARP-l.6, RMS CH FAILURE, FHl, Version 64.0 4.18 IF R-18 alarms with high liquid effluent activity possible, THEN verify any liquid waste release is secured and refer to FNP-I-S0P-50, LIQUID WASTE PROCESSING SYSTEM for potential problems with the liquid waste system. Page: 125 of 200 12/1412009 Previous NRC exam history if any: 068K6.10 068 Liquid Radwaste System K6 Knowledge of the effect of a loss or malfunction on the following will have on the Liquid Radwaste System: (CFR: 41.7 /45.7) K6.l0 Radiation monitors ............................................... 2.5 2.9 Match justification: The only Radiation monitor for which a loss or malfunction would affect the Liquid Radwaste System is R-18. This question requires knowledge of how a failure of R-18, during a liquid waste release, would affect the Liquid Radwaste system. In order to produce 3 plausible but incorrect distractors, the required actions are part of each choice in the second part. Objective:

5. DEFINE AND EVALUATE the operational implications of normal/abnormal plant or equipment conditions associated with the safe operation of the Radiation Monitoring System components and equipment, to include the following (OPS-40305A07):

Normal control methods
Abnormal and Emergency Control Methods
Automatic actuation
Protective isolations
Protective interlocks
Actions needed to mitigate the consequence of the abnormality Page: 126 of 200 1211412009 07/02/09 06:30:42 FNP-1-ARP-1.6 LOCATION FH1 RADIATION MONITOR REFERENCE TABLE (cont) RE LOCATION TYPE DETECTOR FUNCTION ACTIONS R-12* Containment Atmosphere Gas G-M(W) Perform Step (AB 121') 4.12 R-13 Waste Gas Compressor Gas G-M(W) Perform Step Suction (AB 100' WGC 4.13 Valve Room) R-14 Plant Vent Stack (AB Roof) Gas G-M(W) Closes HCV-14 Perform Step ODCM 4.14 R-15A Condenser Air Ejector Gas G-M Perform Step ODCM Discharge Header (TB 155') 4.15A R-15B* Condenser Air Ejector Gas G-M (Eberline)

Perform Step (Intermediate Range) (TB 4.15B 189') R-15C* Condenser Air Ejector Gas Ion Chamber Perform Step (High Range) (TB 189') (Eberline) 4.15B R-17A Component Cooling Water Liquid Scinto (W) Closes CCW Perform Step (CCW Hx Room) surge tank vent 4.17 (RCV-3028) R-17B Component Cooling Water Liquid Scinto Closes CCW Perform ?-P--j::J. (CCW Hx Room) surge tank vent 4.17 0 -1) P. . (RCV-iO?R) ,.--R-18 Waste Monitor Tank Pump Liquid Scinto (W) ( Closes RCV-18 Perform Step ODCM Discharge (AB 121' at the I"--4.18 gP-(i-Batching Funnel) R-19 Steam Generator Liquid Scinto (W) Isolates sample Perform Step , Blowdown/Sample (AB lines 3328, 4.19 139') 3329,3330 R-20A Service Water from Liquid Scinto (W) Perform Step Containment Coolers A and 4.20 -B (AB 121' BTRS Chiller -Room) R-20B Service Water from Liquid Scinto (W) Perform Step Containment Coolers C and 4.20 D (AB 121') *Technical Specification related Page 5 of 12 Version 61.0 07/02109 06:30:42 FNP-I-ARP-l.6 LOCATION FHl P/C9-'1I ;;lki f4 ',,/ OPERATOR ACTION (cont} . I vv..crr 4.18 IF R-18 alarms with high liquid effluent activity possible, any liquid waste release is secured and refer to FNP-I-S0P-50, LIQUID WASTE PROCESSING SYSTEM for potential problems with the liquid waste system. 4.19 IF R-19_alarms AND remains above the alarm setpoint (not a momentary spike), THEN notify the Counting Room to immediately sample the SGs per FNP-O-CCP-31, LEAK RATE DETERMINATION, to determine the leak rate. Refer to FNP-I-S0P-45.0, RADIATION MONITORING SYSTEM for guidance in sampling steam generators with R-19 in alarm. 4.20 R-20A AND R-20B would not normally be expected to indicate high radioactivity since SW pressure is higher than the opposite side of the 'coolers' upstream ofR-20A and R-20B. Request Counting Room to sample the SW effluent. IF high activity is confirmed, THEN investigate for a possible cross system connection to the SW system. {CMT 0005153} 4.21 IF R-21 alarms, THEN implement FNP-O-EIP-9.0, EMERGENCY CLASSIFICATION AND ACTIONS. {CMTs 0008751, 0008755}. 4.22 IF R-22 alarms, THEN implement FNP-O-EIP-9.0, EMERGENCY CLASSIFICATION AND ACTIONS. {CMTs 0008751, 0008755}. 4.23 IF R-23A OR R-23B alarms AND remains above the alarm setpoint (not a momentary spike), THEN perform the following: 4.23.1 Notify the Counting Room to immediately sample the SGs per FNP-O-CCP-31, LEAK RATE DETERMINATION, to determine the leak rate 4.23.2 Contact the RAD man to verify blowdown secured. 4.24 IF R-26A OR R-26B alarms, THEN refer to FNP-I-S0P-50, LIQUID WASTE PROCESSING SYSTEM for potential problems with the liquid waste system.

References:

A-I77100, Sh. 306; U-260841; D-181751; D-181752; D-181753; FSAR, Section 11.4 Page 12 of 12 Version 61.0 ) 07/02/09 06:30:42 SETPOINT: Not Applicable ORIGIN: Any of the below listed Area, Process or Gaseous and Particulate Monitors: ROl, R02, R03, R04, R05, R06, R07, R08, RIO, Rll, R12, R13, R14, R15, R17A RI7B, R18, R19, R20A, R20B, R21, R22, R23A, R23B,R26A,orR26B. PROBABLE CAUSE H2 FNP-I-ARP-l.6 LOCATION FH2 RMS CHFAILURE

1. Loss of input signal to any of the above listed Radiation Detection Channels.

2. Loss of Power to a Channel. 3. Radiation Monitoring System testing in progress.

AUTOMATIC ACTION 1. The radiation monitors fail to a "High Radiation" condition on loss of instrument and/or control power that will result in actuation of associated automatic functions. Refer to annunciator FHl for automatic actions. OPERATOR ACTION I NOTE: Low Alarm Light "on" indicates failure. 1. Check indications on radiation monitoring system console and determine which radiation monitor channel indicates a failure. 2. Notify chemistry and health physics personnel.

3. Notify Instrument Service Personnel to: A. Investigate the failure. B. Make repairs as necessary.

4. Return the Radiation Monitor System Channel to service, in accordance with FNP-I-SOP-45.0, RADIATION MONITORING SYSTEM, as soon as possible.

5. Refer to the Technical Requirements Manual section on Radiation Monitoring Instrumentation.

References:

A-177100, Sh. 307; U-260841; D-181751; D-181752; D-181753; FSAR, Section 11.4 Page 1 of 1 Version 61.0

1. LIQ SD WAST-52106A05 003IHLTIICIA 4.6/4.4/APE059AG2.4.49INESII 068K6.10 A #2 Waste Monitor Tank release to the environment is in progress in accordance with a Liquid release permit and SOP-50.1, Appendix 2, Waste Monitor Tank #2 Release to the Environment.

Annunciator FH2, RMS CH FAILURE, alarms.
R-18, LlQ WASTE DISCH, is indicating normal on the Radiation Monitoring system console and on the recorder for R-18, RR0200.
The HIGH Alarm and LOW Alarm red lights are illuminated.
The control power fuse is found to be illuminated on R-18. When called, the Radside SO reports that Waste Monitor Tank Pump #2 discharge flow transmitter, FT-1085A, indicates 35 gpm. Which one of the following describes the actions required lAW SOP-50.1, Liquid Waste Processing System Liquid Waste Release From Waste Monitor Tank? Direct the Radside SO to __________

_ A. fail air to RCV-18, WMT Disch to Environment OR close the manual discharge valve to the environment. B. fail air to RCV-18, WMT Disch to Environment OR using the manual handwheel, close RCV-18, WMT Disch to Environment. Cy close WMT Disch To Environment, RCV-18 at the LWP OR close the manual discharge valve to the environment. D. close WMT Disch To Environment, RCV-18 at the LWP OR using the manual handwheel, close RCV-18, WMT Disch to Environment. Page: 10f2 10/26/2009 APE059G2.4.49 059 Accidental Liquid Radwaste Release 2.4 Emergency Procedures / Plan 2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls. (CFR: 41.10/43.2/45.6) IMPORTANCE RO 4.6 SRO 4.4 A. Incorrect -Plausible since failing air to RCV-18 could stop the release if the problem was other than mechanical binding, but this is not the procedurally correct way to secure the release lAW SOP-50.1. RCV-18 fails closed on a loss of air. Also, if RCV 18 did not close on the auto close signal, it may be mechanically bound and not shut when air is failed. The second part is correct. B. Incorrect -First part is incorrect (see A). Second part is incorrect due to not being in the procedure and the valve does not have a manual handwheel. Plausible since many AOVs have manual handwheels and would be used to close their respective valve if necessary. C. Correct -SOP-50.1 states that if the discharge is in progress and R-18 becomes inoperable the discharge is to be immediately stopped and the Shift Support Supervisor notified. The manual discharge valve will stop the release since RCV-18 did not close and is the procedurally correct method to secure the release at the end of SOP-50.1. RCV-18 Handswitch at the Liquid Waste Panel (LWP) is also directed to be taken to the closed position in SOP-50.1 and may also isolate the release. D. Incorrect -The first part is correct (see C). The second part is incorrect (see B). 2008 NRC exam Technical

Reference:

SOP-50.1 Ver. 60.0, AOP-6 Ver. 31, LIQUID WASTE PERMIT, ARP-1.6 Ver. 58, FH1 AND FH2 Comments: This question tests the Immediate Actions of a procedure that is used by both the systems operator and CRO to perform a release. Since this deals with the INOPERABLE side of R-18 from the Control room and the required actions should this occur at an RO level for a liquid release, it meets the KA. Page: 2 of2 10126/2009

48. 069AA2.01 OOllFNP BANKJROIMEM 3.7/4.3/Y 2007/N/2/CVRlY Which one of the following conditions represents a loss of containment integrity and would cause entry into Tech Spec 3.6.1, Containment?

A. Mode 3 and one of the Personnel Airlock doors will not close. B!"" Mode 4 and Integrated Leak Rate test determines that leakage is not within limits. C. Mode 5 and it is discovered that the Phase 'B' isolation valve for CCW to the RCPs, will not close. D. Mode 6 and the Equipment Hatch is held in place by 4 bolts ONLY. A -incorrect. Both doors inop would be a loss of Containment Integrity, this is just an inop of one of the doors in the Personnel Airlock. Plausible because one of two series valves at a containment penetration makes containment integrity LCO not met. B -correct. Surveillance requires ILRT to be within limits for Containment Integrity to be set. C -incorrect. because Containment Integrity is not required in Mode 5, plausible because the valve is part of a containment penetration that would affect integrity in modes 1-4. o -incorrect. 4 bolts meets the minimum requirement for Containment Closure in Mode 6, but not containment integrity in the modes that containment integrity is required. TS 3.6.1 Previous NRC exam history if any: 2007 FNP NRC EXAM, this is the only question in the FNP bank tied to this k/a. 069AA2.01 069 Loss of Containment Integrity AA2. Ability to determine and interpret the following as they apply to the Loss of Containment Integrity: (CFR: 43.5/45.13) AA2.0 1 Loss of containment integrity ...................................... 3.7 4.3 Match justification: This question requires knowledge of the ability to determine IF Ctmt integrity is lost or met in different modes lAW Tech Specs. Mode applicability (1-4) & one hour or less tech specs (one or more air locks with one door inoperable) are RO level Knowledge. Objective: OPS-52102A-1 Page: 127 of 200 12/14/2009 3.6 CONTAINMENT SYSTEMS 3.6.1 Containment Containment 3.6.1 LCO 3.6.1 Containment shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, a@ ACTIONS CONDITION A. Structural integrity of the containment not conforming to the requirements of SR 3.6.1.2. (j' B. Containment inoperable for reasons other t,/ Condition A. C. Required Action and associated Completion Time not met. Farley Units 1 and 2 A.1 B.1 C.1 Atill. C.2 REQUIRED ACTION COMPLETION TIME Restore the structural 24 hours integrity to within limits. Restore containment to 1 hour OPERABLE status. Be in MODE 3. 6 hours Be in MODE 5. 36 hours 3.6.1-1 Amendment No. 146 (Unit 1) Amendment No. 137 (Unit 2) SURVEILLANCE REQUIREMENTS SR 3.6.1.1 SR 3.6.1.2 SURVEILLANCE Perform required visual examinations and leakage rate testing except for containment air lock testing, in accordance with the Containment Leakage Rate Testing Program . .--, Verify containment structural integrity in accordance with the Containment Tendon Surveillance Program. Containment 3.6.1 FREQUENCY In accordance with the Containment Tendon Surveillance Program Farley Units 1 and 2 3.6.1-2 Amendment No. 146 (Unit 1) Amendment No. 137 (Unit 2) 3.6 CONTAINMENT SYSTEMS Containment Air Locks 3.6.2 3.6.2 Containment Air Locks T .:;) f} ) Cz;> f W Lee Two containment air locks shall be OPERABEl APPLICABILlT: MODES 1, 2,@jand 4. ACTIONS --------------------

NOT E S--------------------------------------------------------------

1. xit is permissible to perform repairs on the affected air lock components.

2. on ition entry is allowed for each air lock. 3. . able Conditions and Required Actions of LCO 3.6.1, "Containment," when air ge results in exceeding the overall containment leakage rate. -----------------------------------------------------------------------------------------------------------------------------

A. One or more containment air locks with one containment air lock door inoperable. Farley Units 1 and 2 REQUIRED ACTION ----------------N 0 TES-------------------- Required Actions A.1, A.2, and A.3 are not applicable if both doors in the same air lock are inoperable and Condition C is entered. 2. Entry and exit is permissible for 7 days under administrative controls if both air locks are inoperable. COMPLETION TIME ( continued) 3.6.2-1 Amendment No. 146 (Unit 1) Amendment No. 137 (Unit 2) 3.6 CONTAINMENT SYSTEMS 3.6.3 Containment Isolation Valves Containment Isolation Valves 3.6.3 LCO 3.6.3 Each containment isolation valve shall be OPERABLE. The 8-inch containment mini-purge supply and exhaust isolation valves may be open for safety-related reasons. APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS ----------------------------------------------------------- NOT E S--------------------------------------------------------

1. Penetration flow path(s) except for 48-inch purge valve flow paths may be unisolated intermittently under administrative controls.

2. Separate Condition entry is allowed for each penetration flow path. 3. Enter applicable Conditions and Required Actions for systems made inoperable by containment isolation valves. 4. Enter applicable Conditions and Required Actions of LCO 3.6.1, "Containment," when isolation valve leakage results in exceeding the overall containment leakage rate acceptance criteria.

Farley Units 1 and 2 3.6.3-1 Amendment No. 146 (Unit 1) Amendment No. 137 (Unit 2) Containment Penetrations 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Containment Penetrations LCO 3.9.3 y )" Ctri The containment penetrations shall be in the following status: a. b. One door in each air lock is capable of being closed; and c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere either: 1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or 2. capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System. APPLICABILITY: During CORE ALTERATIONS, During movement of irradiated fuel assemblies within containment. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more containment A.1 Suspend CORE AL TERATIONS. Immediately penetrations not in required status. Farley Units 1 and 2 AND A.2 Suspend movement of Immediately irradiated fuel assemblies within containment. 3.9.3-1 Amendment No. 178 (Unit 1) Amendment No. 171 (Unit 2)

49. 071A4.01 OOllNEW/ROIMEM 2.7/2.21N1N13/GTOICVRIY Unit 1 is 60% power at EOL, and the following conditions exist: * #1 RHT is On Service and level is 10%.

LK-112, L TDN TO VCT FLOW, has been adjusted to maintain the VCT at 40% level in AUTO.
A manual makeup of 400 gallons at a rate of 40 gpm has been set up as the unit ramps up in power. * #2 RHT is Off Service and level is 50%. * #2 RHT gas space under the bladder is being educted lAW SOP-2.4, Chemical And Volume Control System Boron Recycle System, using the step entitled "#1 (#2, #3) RHT venting with RHT GAS SAMPLE PANEL." Which one of the following describes:

1) the indication of LCV115A, VCT HI LVL DIVERT VLV, in response to the manual makeup and 2) the indication that the educting of #2 RHT is complete or almost complete lAW SOP-2.4? A'! 1) LCV115Awill indicate RED (VCT) AND WHITE (HU TANK) lights LIT. 2) Gas panel annunciator

23, RHT EDUCTOR LO PRESS, comes into alarm when educting is almost complete.

B. 1) LCV115Awill indicate RED (VCT) AND WHITE (HU TANK) lights LIT. 2) PCV-251, RHT Eductor Suction Line Pressure Control valve, indication will change from red light LIT to green light LIT when educting is complete. C. 1) LCV115A will indicate RED (VCT) light ONLY LIT. 2) Gas panel annunciator

23, RHT EDUCTOR LO PRESS, comes into alarm when educting is almost complete.

D. 1) LCV115A will indicate RED (VCT) light ONLY LIT. 2) PCV-251, RHT Eductor Suction Line Pressure Control valve, indication will change from red light LIT to green light LIT when educting is complete. VCT = 15 gal/%. If VCT level at 20%, then a 400 gallon add will result in level rise to 46%. An EOL ramp up, it is operationally valid to add 400 gallons for Xe buildup. On the LCV115A lights on the MCB the following is written: RED (VCT) AND WHITE (HU TANK) This is the reason this is provided in the distracter. A -Correct. LCV 115A is a three way valve that has only a red and a white light indicated on the MCB, and unlike most other valve indications of the MCB, it has no Page: 128 of200 12/14/2009 green light. The red light is on when the valve is not fully diverted to the RHT and the white light is on when the valve is not fully aligned to the VCT. The red light indicates at least some flow going to the VCT, and the white light indicates at least some flow going to the RHT. When the valve is in mid position, both lights are on, such as is the case with a 40 gpm makeup. Verified on the simulator laptop computer. The second part is correct per 4.9.14 and associated NOTE of SOP-2.4. This alarm indicates 7" vacuum, and is an indication that educting is almost complete. Educting must be secured by 20" vacuum. B -Incorrect. The first part is correct (see A). The second part incorrect, since PCV-2S1, IF it were open in this lineup, would close at 7" vacuum automatically, but the educting can continue to a higher vacuum than 7" in this educting lineup. Plausible, since PCV-2S1 would automatically close at 7" and secure the educting if the installed piping for educting and installed valve (PCV-2S1) was used for the educting flowpath (per Step 4.8.7 of SOP-2.4). However, in the educting flowpath specified in the stem, using a portable vacuum pump and discharging directly to the plant vent, PCV-2S1 is not in the flowpath, and allowance is made by the procedure to go to a vacuum of 20" vice 7". C -Incorrect. The first part is incorrect (see A). The red light only would be lit if all water was flowing to the VCT. Plausible, since this choice would be chosen if confusion existed as to which valve position the red indicates (normally open on other valve indications). Also, the white light may not be understood. It may be thought to be a full divert indication instead of a partial OR full divert indication. This is a non-standard light indication arrangement, since most valves are not three way, and have red and green lights for open and close indications. For example, the TCV143 hi letdown temperature divert valve HS white light is for the VCT position and is next to the LCV11SA HS, which has the white light for the RHT position and the red light for the VCT position. The second part is correct (see A). 0-Incorrect. The first part is incorrect (see C). The second part is incorrect (see B). A-181009, CVCSfHHSIIACCUMULATORIRMWS 3.14.1 Basic Functions Valve LCV -lISA controls the amount of letdown flow diverted to the RHTs on high VCT level. Valves LCV-IISB,C,D and E are actuated to isolate the VCT on low level and provide a suction source for the charging pumps from the R WST. 3.14.2 Functional Requirements Modulate Divert (LT-112) --This setpoint shall start diversion ofletdown to the RHTs via valve LCV -lISA. The valve shall modulate open as required, based on maintaining the level at the modulate divert setpoint. (References 6.2.1, 6.2.9, 6.2.58, 6.3.32, and 6.3.13) FNP-1-S0P-2.5, RCS Chemical Addition, VCT Gas Control, And Demineralizer Operation, Version 67.0 Appendix 2 Flushing Cation and Mixed Bed Demineralizers to the RHT's 4.2.5 Commence blended or batch makeup to the VCT equal to desired RCS boron concentration per FNP-I-SOP-2.3 CHEMICAL AND VOLUME CONTROL SYSTEM REACTOR MAKEUP CONTROL SYSTEM. Page: 129 of 200 12114/2009 FNP-I-ARP-13.1, Boron Recycle Processing Panel, Version 8.0 Annunciator window 23, Recycle Holdup Tank Eductor Lo Press SOP-2.4, Chemical And Volume Control System Boron Recycle System, Version 55.0 4.1 CAUTION: RHT may overflow if level is allowed to exceed 50% without venting RHT. 4.1.1.10 Monitor RHT level as follows: 1. Frequently check indicated level for expected level rise. NOTES:

On service RHT is normally swapped prior to exceeding 50% level.
SS approval required to exceed 50% level. 2. IF intentionally filling an RHT >50% THEN, verify RHT has been educted within the last 30 days prior to exceeding 50%. 3. WHEN transferring water to an RHT with indicated level> 50% THEN, check bladder pressure approximately every 30 minutes. (Ref. OR 2-96-329)

4. IF bladder pressure>

0.5 psig THEN, educt (vent) RHT using desired section of this procedure. 4.8 #1 (#2, #3) RHT venting with WASTE GAS SYSTEM. 4.9 #1 (#2 #3) RHT venting with RHT GAS SAMPLE PANEL. 4.12 #1 (#2, #3) RHT pressure check for gas buildup under bladder using temporary gage or installed instrumentation. Previous NRC exam history if any: none 071A4.01 071 Waste Gas Disposal System A4 Ability to manually operate and/or monitor in the control room: (CFR: 41.7 145.5 to 45.8) A4.01 Valve to put the holdup tank into service; indications of valve positions and tank pressure .2.7* 2.2* Match justification: The Holdup tank at FNP is the Recycle Holdup Tank (RHT), which is filled (put into service) when LCV-115A, Letdown divert valve to the RHT, shifts to the divert position. LCV-115A position is indicated in the Control Room. This question requires knowledge of the how this valve indicates the position during a continuous dilution evolution. The tie to the Waste Gas system is that RHT pressure can build up under the bladder from gasses coming out of solution, and the question requires knowledge of how to monitor in the control room while operating the holdup tank, and the gas system during educting the holdup tank, at the given water level to educt the gas space under the bladder to mitigate the effects of the pressure accumulation (the actual RHT pressure guage is local in the plant, and not remote in the control room). An RHT pressure alarm is used for indication of when to secure the educting, and it indicates both locally at the Gas panel and at a common Gas and Liquid Waste alarm in the control room. Applying this k/a to FNP is challenging, and making it a discriminating question that isn't trivia is also challenging, but this has been accomplished in this question. Page: 130 of 200 12/14/2009 Discussed with lead examiner 9-21-09 that the only holdup tank at FNP is the Recycle Holdup Tank that received liquid RCS water, and the only connection between the holdup tank and the Waste gas system is when educting the gas space under the bladder (required at FNP when the tank gets to 50% level). The only Control Room indications are for the valve which diverts letdown flow to the RHT placing the holdup tank in service} and the common MCB alarm which alarms when the Educting is almost complete. Diverting flow to the RHT will, as RHT level rises to 50%, require educting with the waste gas system OR the new method of using a portable vacuum pump discharging to the Plant Vent stack. The original method of educting is not normally used (educting with the Waste Gas Compressors to a waste gas decay tank), although it is still installed, trained on, and in the procedure. The old method could still be used at any time, but a portable vacuum pump is normally used to educt the tank directly to the Plant Vent Stack. Objective:

7. DEFINE AND EVALUATE the operational implications of normal! abnormal plant or equipment conditions associated with the safe operation of the Waste Gas System components and equipment, to include the following (OPS-40303B07):

Normal control methods Abnormal and Emergency Control Methods Automatic actuation including setpoint (example SI, Phase A, Phase B, MSLIAS, LOSP, SG level) Protective isolations such as high flow, low pressure, low level including setpoint Protective interlocks Actions needed to mitigate the consequence of the abnormality

2. RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the components associated with the Chemical and Volume Control System, to include the components found on Figure 3, Chemical and Volume Control System and Figure 4, RCP-Seal Injection System (OPS-40301F02).

Page: 131 of 200 12/14/2009 FNP Units I & 2 3.13.3 3.13.4 3.13.5 3.14 VeT LEVEL LICA-112 LIC-115 3.14.1 WESTINGHOUSE PROPRIETARY CLASS 2 CVCS/HHSII ACCUMULA TORlRMWS A-181009 temperature, range, orifice I.D and readout location. (References 6.2.9, 6.5.29.v, 6.5.29.x, 6.5.29.aa and 6.5.73) Equipment Qualification Requirements 3.13.3.1 The electronic transmitters shall be environmentally qualified as detailed in references 6.7.26, 6.7.27, 6.7.29.h and 6.7.29.i. Maintenance specified in the EQ package shall be performed to maintain qualified life. (References 6.7.23, 6.7.26,6.7.27, 6.7.29.h, 6.7.29.i and 6.7.28) Interface Requirements 3.13.4.1 The channel must be powered from a 120 Volt-AC regulated instrument power system. This regulated AC power shall be 118 volts +/- 2%, 60 cps +/- 2% nominal, 3% maximum harmonic distortion (normal), 5% maximum. (Reference 6.5.21) Equipment Protection Features 3.13.5.1 The differential pressure device shall be capable of withstanding a sustained input pressure of 1.5 times system design pressure. This applied pressure shall not damage the instrument so that it can perform its function. (Reference 6.5.21) Basic Functions 3.14.1.1 These level instruments shall provide measurement ofVCT level and shall supply signals for control and actuation of valves LCV -115A, B, C, D and E. Valve LCV -115A controls the amount of letdown flow diverted to the RHTs on high VCT level. Valves LCV-115B,C,D and E are actuated to isolate the VCT on low level and provide a suction source for the charging pumps from the RWST. The instruments shall alert the control room operator 3-31 Ver.31.0 FNP Units 1 & 2 3.14.2 WESTINGHOUSE PROPRIETARY CLASS 2 CVCSIHHSII ACCUMULATORlRMWS A-181009 3.14.1.2 of high-and low-level conditions in the VCT and when the Reactor Makeup Control System controls are not positioned properly to support auto-makeup to the VCT. (References 6.2.1, 6.2.56, 6.2.57, 6.2.9 and 6.4.1) Control Room indication ofVCT level is required by RG 1.97 for post accident monitoring. (References 6.7.24, 6.7.25 and 6.7.23) Functional Requirements 3.14.2.1 3.14.2.2 Level indication shall be provided on the main control board for LI -115. This provides the operators with indication for normal operations and is also used to satisfy post accident monitoring (RG 1.97) requirements. The signal from LI-112 shall be indicated locally. (References 6.2.1, 6.2.9, 6.4.1,6.7.24,6.7.25 and 6.7.23) The instruments shall support the following level setpoint requirements. (Refer to References 6.2.7 and 6.2.8 for the specific alarm and control setpoint): Emergency Makeup Start RWST --Upon coincident (2/2 logic) signals from both level instruments, valves LCV-115B and D (QV336A1B) from the RWST shall open, and valves LCV-115C and E (QV376A/B) from the VCT shall close to maintain suction to the charging pumps. Should one of these signals fail to actuate, the operator still has at least 2.5 -minutes time period before the potential suction loss to the charging pumps. The 2.5-minute time period assumes 120-gpm letdown being diverted to the holdup tank and no makeup supplied by the Reactor Makeup Control System. (References 6.2.1, 6.2.9, 6.2.58, 6.3.12, 6.3.13 and 6.2.56) Emergency Makeup Stop R WST --This setpoint provides for the clear signal of the emergency makeup start and allows the operator to reposition the RWST and VCT isolation valves to their normal position. (References 6.3.12,6.3.13 and 6.2.58) Low Level Alarm (L T -112 and L T -115) --A visual and audible alarm shall be provided on the main control board, warning the operator of a low-level condition. This setpoint 3-32 Rev. 18 WESTINGHOUSE PROPRIETARY CLASS 2 FNP Units 1 & 2 CVCS/HHSII ACCUMULATORlRMWS A-18l009 NOT Stated in FSD, Butverifiedon simulator: White light indicates an Auto open signal (modulate OR full. divert), and is. to the right of the handswitchlight array inthe position of the green light of other valves; The red light .indicates the thevalye is to the VCT position RedlWhite indicates differently than other valves (there is no green): Red = at least partially to the VCT position White = at least partially to the RHT position is below the setpoint for automatic initiation of makeup to the VCT. This warns the operator that the automatic makeup function has failed to actuate or that the system is not causing the VCT level to rise. (References 6.2.1, 6.2.9, 6.2.58,6.3.12 and 6.3.13) Auto Makeup Start (L T -115) --This setpoint shall start auto makeup to the VCT from the Reactor Makeup Control System and activate and alarm if the Reactor Makeup Control System is not set in its auto makeup mode. This setpoint should be activated when the VCT level is approaching the bottom of the tank but above the three level setpoints discussed above. The setpoint was arbitrarily set 3-inches above the low level alarm. (References 6.2.1, 6.2.9,6.2.58,6.3.12 and 6.3.13) shall be provided on the main control board, warning the operator of a high-level condition. This setpoint is above the setpoint for modulate divert of makeup to the VCT. This warns the operator that the modulate divert function has failed to actuate or that the system is not causing the VCT level to drop. The high alarm also warns the operator that the level is approaching the maximum level recommended for degassing operations. (References 6.2.1, 6.2.9,6.2.58,6.3.12 and 6.3.13) Fully Divert (L T -115) --This setpoint shall cause valve LCV -liSA to fully divert letdown to the RHTs. The full divert function is provided as a back-up in the event that the modulate divert function (performed by L T -112) should fail. The primary restricting consideration here is to prevent 3-33 Rev. 18 FNP Units 1 & 2 3.14.3 3.14.4 3.14.5 WESTINGHOUSE PROPRIETARY CLASS 2 CVCS/HHSII ACCUMULATORlRMWS A-181009 3.14.2.3 pressurizing the VCT over its maximum normal operating pressure of 65 psig and risk lifting the safety valve. (References 6.2.1, 6.2.9, 6.2.58, 6.3.12, and 6.3.13) Refer to Table T -11 for instrumentation design requirements including instrumentation type, design pressure, temperature, range and readout location. (References 6.2.9 and 6.5.29.e) Equipment Qualification Requirements 3.14.3.1 The electronic transmitter (L T -115) shall be environmentally qualified as detailed in references 6.7.26, 6.7.27, 6.7.29.h and 6.7.29.i. Maintenance specified in the EQ package shall be performed to maintain qualified life. (References 6.7.23,6.7.26,6.7.27, 6.7.29.h, 6.7.29.i and 6.7.28) Interface Requirements 3.14.4.1 The channel must be powered from a 120 Volt-AC regulated instrument power system. This regulated AC power shall be 118 volts +/- 2%, 60 cps +/- 2% nominal, 3% maximum harmonic distortion (normal), 5% maximum. (Reference 6.5.21) Equipment Protection Features 3.14.5.1 The differential pressure device shall be capable of withstanding a sustained input pressure of 1.5 times system design pressure. This applied pressure shall not damage the instrument so that it can perform its function. (Reference 6.5.21) 3.15 ACCUMULATOR TANK PRESSURE PIA-921 (Tank 1) PIA-925 (Tank 2) PIA-929 (Tank 3) PIA-923 (Tank 1) PIA-927 (Tank 2) PIA-931 (Tank 3) 3.15.1 Basic Functions 3.15.1.1 These pressure instruments shall be utilized by the operator to help set and maintain the gas overpressure in the accumulators within the Technical Specification limits 3-34 Rev. 18 05112/09 12:36:00 1.0 Purpose F ARLEY NUCLEAR PLANT UNIT 1 SYSTEM OPERATING PROCEDURE SOP-2.4 CHEMICAL AND VOLUME CONTROL SYSTEM BORON RECYCLE SYSTEM FNP-1-S0P-2.4 This procedure provides Initial Conditions, Precautions and Limitations, and Instructions necessary for the operation of the Boron Recycle System. Instructions are included in the following sections: 4.1 #1 (#2, #3) RHT operation. 4.1.1 Swapping In Service RHT's. 4.2 Recycle evaporator feed pump operation. 4.3 Placing RHT on Big Recirc or Transferring between RHTs via the recycle evaporator feed demineralizers and filters. 4.4 Placing RHT on Big Recirc or Transferring between RHTs with the recycle evaporator feed demineralizers and filters B YP AS SED. 4.5 #1 (#2, #3) RHT discharge to charging pump suctions. 4.6 #1 (#2, #3) RHT discharge to SFPCS transfer canal. 4.7 #1 (#2, #3) RHT discharge to waste evaporator. (Deleted by Version 41.0) 4.8 #1 (#2, #3) RHT venting with WASTE GAS SYSTEM. #1 (#2 #3) RHT venting with RHT GAS 4.10 Draining RHTs to WHT or FDT. 4.11 Transferring between RHTs using recycle evaporator feed pump mini flow line. 4.12 #1 (#2, #3) RHT pressure check for gas buildup under bladder using temporary gage or installed instrumentation. Appendix 1 Transfer of Unit-1 RHTs to Unit-1 RWST Appendix 2 Transfer of Unit 1 RHTs to Unit 2 Version 54.0 05112109 12:36:00 FNP-I-S0P-2.4 4.8 #1(#2, #3) RHT venting with WASTE GAS SYSTEM. (CMT 0005260) NOTE: Indicate completion of applicable (*) steps by initialing on procedure sign-off list FNP-l-SOP-2.4B.

4.8.1 Transfer the water from the RHT which is to be vented per section 4.4 until level is between 50% and 5%. Level may be left higher than 50% with Shift Supervisor's permission.

4.8.1.1 4.8.1.2 4.8.1.3 IF steps 4.8.1.2 and 4.8.1.3 are required by the Shift Supervisor, THEN notify Maintenance to remove manway on #1 (#2, #3) RHT. Have Health Physics check above the bladder for combustible gas concentration to detect bladder leakage. Operations personnel determine the size of the gas bubble. NOTE: The waste gas system must be aligned to the low pressure mode for operations with the eductor. *4.8.2 Align off service waste gas compressor to eductor as follows: 4.8.2.1 Notify Health Physics Foreman prior to educting RHTs due to dose rate changes on eduction piping. 4.8.2.2 Verify RMW and CCW aligned to off service compressor. 4.8.2.3 Verify open compressor suction valve I-GWD-V-7907A(B) (QIG22V06IA [B]). 4.8.2.4 Close compressor discharge valve to recombiners I-GWD-V-7910A (B) (QIG22V064A [B]). 4.8.2.5 Open compressor discharge to eductor I-GWD-V-7911A(B) (QIG22VI97A [B]). 4.8.2.6 Open eductor return isolation I-GWD-V -7807 (QIG22V008). 4.8.2.7 Start lA (IB) waste gas compressor. (Step 4.8 continued on next page) Version 54.0 05/12/09 12:36:00 FNP-I-S0P-2.4 CAUTION: In event*of a significant diaphragm leak, reactor coolant system draining, or spent fuel pit draining, align the waste gas compressor to 8: shutdown tank per FNP-I-SOP-51.0,WASTEGAS SYSTEM, instead of the normal path to a gas decay tank. *4.8.3 Close or verify closed the following valves: * #1 (#2, #3) RHT MAIN INLET ISO l-CVC-V-8554A (B,C) (QIE21 V284A [B,C]). * #1(#2, #3) RHT MINIFLOW ISO valve l-CVC-V-8556A (B,C) (QIE21 V309A[B,C]).

#IRHT EDUCTOR SUCT QIE21V314A.
#2 RHT EDUCTOR SUCT QIE21V314B.
#3 RHT EDUCTOR SUCT QIE21V314C.
4.8.4 Verify locked closed RHT equipment drains and valve leakoffs inlet line isolation l-CVC-V-8557A(B,C) (QIE21V311A[B,CD. (Master Z Key) *4.8.5 Open or verify open the following valves: * #1(#2, #3) RHT EDUCTOR SUCT QIE21 V314A(B,C) for the RHT to be vented.
RHT eductor suction line pressure control l-CVC-PCV

-251 (Q 1 E21 V366).

RHT EDUCTOR SUCT ISO valve QIE21V319.
RHT eductor suct sample bypass l-CVC-V-8638 (QIE21V321). (Step 4.8 continued on next page) Version 54.0 05112/09 12:36:00 FNP-1-S0P-2.4 NOTE: Contact HP when venting RHTs to take periodic air samples of the AUX building due to the potential of creating airborne areas when venting hot RHT. 4.9 #1 (#2, #3) RHT venting with RHT GAS SAMPLE PANEL. NOTE: Indicate completion of applicable

(*) steps by initialing on procedure sign-off list FNP-l-SOP-2.4B.

4.9.1 Transfer the water from the RHT which is to be vented per section 4.4 until level is between 50% and 5%. Level may be left higher than 50% with Shift Supervisor's permission.

4.9.1.1 IF steps 4.9.1.2 and 4.9.1.3 are required by the Shift Supervisor, THEN have Maintenance remove the man way on #1 (#2, #3) RHT. *4.9.1.2 4.9.1.3 Have Health Physics check above the bladder for combustible gas concentration to detect bladder leakage. Operations personnel determine the size of the gas bubble. 4.9.2 Collect the required equipment. 4.9.2.1 4.9.2.2 4.9.2.3 4.9.2.4 Vacuum pump with flow meter suitable for monitoring flow in the range of 0-3 cfm (85 liters/min). (Health Physics air sampler) 50' electrical extension cord with 3 prong plug or adapter. 50' tygon hose. Suction hose, approximately 10' poly flow or thick walled tygon hose. *4.9.3 Verify closed #1, #2, and #3 RHT sample line isolation valves (Aux *4.9.4 *4.9.5 Bldg, 121' in hallway outside RHT room inside SAMPLE PANEL):

N1E21 V324A
N1E21V324B
N1E21V324C Verify closed RHT SAMPLE LINE ISO valve N1E21 V325B. (Aux Bldg, 121' in hallway outside RHT room inside SAMPLE PANEL) Connect vacuum pump suction hose to RHT SAMPLE LINE ISO valve N1E21 V325B. (Aux Bldg, 121' in hallway outside RHT room inside SAMPLE PANEL) (Step 4.9 continued on next page) Version 54.0 05112/09 12:36:00 FNP-1-S0P-2.4 NOTE: After hose is routed to the HV AC exhaust duct, it should be blue tagged and left hanging for future venting. Ensure hose is securely routed. *4.9.6 *4.9.7 *4.9.8
4.9.9 *4.9.10 *4.9.11 *4.9.12 *4.9.13 Connect a hose to the discharge of the vacuum pump and route to a HV AC exhaust duct. Verify hose is placed to discharge gases directly into exhaust duct. Verify the radwaste ventilation system is in operation.

Check hose connections and verify all connections are suitable to prevent leakage of H2 gas. Slowly open #1 (#2, #3) RHT SAMPLE LINE ISO valve for the RHT to be vented (Aux Bldg, 121' in hallway outside RHT room inside SAMPLE PANEL):

N1E21V324A
N1E21 V324B
N1E21V324C Slowly open RHT SAMPLE LINE ISO valve N1E21 V325B. (Aux Bldg, 121' in hallway outside RHT room inside SAMPLE PANEL) Start vacuum pump. Adjust flow rate to maintain less than 2.0 cfm (55 liters/min). (Step 4.9 continued on next page) Version 54.0 05/12/09 12:36:00 FNP-1-S0P-2.4 CAUTION: Educti.on should not continue past 20" H 2 0 vacuum to prevent loss of overflow water seal. NOTES: The intent of the following step is to lock in vacuum under the bladder prior to securing the vacuum pump to prevent reintroduction of oxygen. Boron Recycle Panel annunciator

23 (RHT EDUCTOR LO PRES comes in at 7" H20 vacuum and should be used as a prompt tha eduction IS iIiiiOst complete.

f required, eduction may be secured prior to reaching Indicated RHT level may rise significantly due to a slight vacuum being drawn, even if no vacuum is indicated on PIS-2S1. A SMALL addition of N2 may be necessary to return indicated level to normal. Monitor RHT level during N2 addition and secure N2 addition when indicated level returns to normal. *4.9.14 *4.9.15 *4.9.16 *4.9.17 WHEN a vacuum is indicated under the bladder OR flow rate drops to zero, THEN perform the following: 4.9.14.1 Close #1(#2, #3) RHT SAMPLE LINE ISO valve for the RHT that was vented (Aux. Bldg. 121' in hallway outside RHT room inside panel):

N1 E21 V324A
N1E21V324B
N1E21V324C 4.9.14.2 Secure the vacuum pump. Close N1E21V325B RHT SAMPLE LINE ISO valve. (Aux Bldg, 121' in hallway outside RHT room inside SAMPLE PANEL) Disconnect vacuum pump, leaving discharge hose to plant ventilation exhaust duct in place and properly secured and blue tagged. Ensure ends of tygon vent hose are properly covered, taped, and secured after use for contamination control, and radiological protection. (Step 4.9 continued on next page) Version 54.0 05/12/09 12:36:00 FNP-1-S0P-2.4 NOTE: *4.9.l8 IF vacuum is indicated OR it is otherwise desired to add N2, THEN notify Chemistry to introduce N2 under the RHT bladder per FNP-1-CCP-667.

Unless further eductions are planned, N2 should only be added to the point of breaking vacuum and should not cause a positive pressure under the bladder. Ifvacuum is indicated, do not continue until the vacuum has been broken by addition of N2 under the bladder. *4.9.l9 4.9.20 *4.9.21 4.9.22 Ensure all equipment is properly stored. Notify the Health Physics Foreman that the RHT venting process is complete. IF required, THEN have Mechanical Maintenance reinstall manway on #1 (#2, #3) RHT. WHEN RHT eduction is complete, THEN ensure an Autolog entry is made for documentation. Version 54.0 so. 072G2.1.27 OOllNEW/RO/MEM 3.9/4.01N1N12/CVRIY Which one of the following states only correct purposes and/or functions of the Area Radiation Monitoring System (ARMS)? A":I *

B. *
C. *
D. *
Provide warning of any radiation hazard that could develop. Provide advance warning of a plant malfunction that could lead to a health hazard or plant damage. Provide warning of any radiation hazard that could develop. Provide control functions to shift ventilation to recirculation in the event of high radioactivity.

Provide control functions to isolate liquid effluent processes in the event of high radioactivity. Provide advance warning of a plant malfunction that could lead to a health hazard or plant damage. Provide control functions to isolate liquid effluent processes in the event of high radioactivity. Provide control functions to shift ventilation to recirculation in the event of high radioactivity. A -Correct. Per FSD: Radiation Monitoring System, A-18101S. The FSD gives the function for the entire RMS system, of which the ARMS is a subset along with PERMS and the Atmosphere Radiation Monitoring System. Each section in the FSD under 3.1, Area Monitors 3-1 describes the functions of each ARMS monitor. Knowing which RMS monitors are ARMS monitors and the different functions of each monitor is required to answer this question correctly. B -Incorrect. The first part is correct (see A). The second part is incorrect. Plausible, since one of the Area radiation monitors stops the ventilation fans (Low Level Rad Waste Building area monitors), but it does not shift to recirc. Other area monitors such as R-1A & R-1 B, CR & TSC respectively, have ventilation that shifts to recirc on high radiation, but due only to R-3SA & 3SB, but not due to any ARMS monitor. C -Incorrect. The first part is incorrect. Plausible, since other parts of the RMS system than the ARMS have this function (PERMS), but not the ARMS. The second part is correct (see A). 0-Incorrect. Both parts are incorrect (see C & B). FSD: Radiation Monitoring System, A-181015 3.1 Area Monitors 3-1 3.1.1 Control Room Area Monitor (RE-0001) 3-1 3.1.2 Technical Support Center Area Monitor (RE-OOO I B) 3 -3 3.1.3 Containment Elevation 155'-0" Area Monitor (RE-0002) 3-4 3.1.4 Radio Chemistry Laboratory Area Monitor (RE-0003) 3-6 Page: 132 of 200 12/1412009 3.1.5 Charging Pump Room Area Monitor (RE-0004) 3-7 3.1.6 Fuel Storage Pool Area Monitor (RE-0005) 3-9 3.1.7 Sampling Room Area Monitor (RE-0006) 3-10 3.1.8 Incore Instrument Area Monitor (RE-0007) 3-12 3.1.9 Drumming Station Area Monitor (RE-0008) 3-13 3.1.10 Sample Panel Room Area Monitor (RE-0009) 3-14 3.1.11 Containment High Range Area Monitor (RE-002 7 A, B) 3 -16 3.1.12 Low Level Radwaste Building Area Monitor (RE-0066A, B, C, D, E, F) 3-18 3.2 Process Liquid Monitors 3-19 3.2.1 Recycle Evaporator Condensate Discharge (RE-00I6) UNIT 1 -SPARED 3-19 3.2.2 Component Cooling Water Pump Suction (RE-00I7 A, B) 3-23 3.2.3 Waste Monitor Tank Discharge Radiation Monitor (RE-0018) 3-26 3.2.4 Steam Generator Blowdown Sample Radiation Monitor (RE-00I9) 3-30 3.2.5 Containment Cooler Service Water Outlet (RE-0020A,B) 3-32 3.2.6 Steam Generator Blowdown (RE-0023A,B) 3-34 3.2.7 Closed Loop Auxiliary Steam System (RE-0026A, B) 3-39 3.2.8 Main Steam Line Nitrogen 16 (N-I6) Monitor (RE-0070A, B, C) 3-41 3.3 Particulate, Iodine, and Gas Monitors 3-43 3.3.1 Penetration Room Filtration Exhaust Particulate Monitor (RE-OOIO) 3-43 3.3.2 Containment Atmosphere Particulate and Noble Gas Monitor (RE-00Il/00I2) 3-46 3.3.3 Waste Gas Processing System Noble Gas Radiation Monitor (RE-0013) 3-50 3.3.4 Plant Vent Stack Noble Gas Radiation Monitor (RE-00I4) 3-53 3.3.5 Steam Jet Air Ejector Exhaust Noble Gas Monitor (RE-I5, B, C) 3-56 3.3.6 Plant Vent Stack Particulate and Noble Gas Monitor (RE-002l/0022) 3-59 3.3.7 Containment Purge Noble Gas Monitor (RE-0024A, B) 3-64 3.3.8 Spent Fuel Pool Ventilation Noble Gas Monitor (RE-0025A, B) 3-68 3.3.9 Steam Jet Air Ejector Exhaust Grab Sampler (RE-0028) 3-72 3.3.10 Plant Vent Stack Particulate/IodinelNoble Gas Monitor (RE-0029B) 3-73 3.3.11 Radwaste Area Ventilation Particulate and Noble Gas Monitor, Elevations 100'-0", 83'-0", and 77'-0" (RE-0030A, B) 3-76 3.3.12 Radwaste Area Ventilation Particulate Monitor, Elevation 121' -0" (RE-003I) 3-79 3.3.13 Radwaste Area Ventilation Particulate Monitor, Elevation 139'-0" (RE-0032) 3-82 3.3.14 Radwaste Area Ventilation Particulate Monitor, Elevation 155'-0" (RE-0033) 3-84 3.3.15 Radwaste Area Ventilation Particulate Monitor, Elevation 155'-0" (Access Control Area) (RE-0034) 3-87 3.3.16 Control Room Ventilation Inlet Noble Gas Monitor (RE-0035A, B) 3-89 3.3.17 Main Steam Safety Relief and TDAFW Pump Exhaust Noble Gas Monitors (RE-0060A, B, C, D) 3-93 3.3.18 Plant Vent Stack Particulate/IodinelNoble Gas Grab Sampler (RE-0029A) 3-96 3.3.19 Containment Post Accident Particulate/IodinelNoble Gas Grab Sampler (RE-0067) 3-97 3.3.20 Plant Vent Stack Particulate/IodinelNoble Gas Grab Sampler (RE-0068) 3-101 3.3.21 Containment Purge Exhaust Particulate/IodinelNoble Gas Grab Sampler (RE-0069) 3-104 1.1 SYSTEM OVERVIEW The RMS is designed to perform three basic functions:

Provide warning of any radiation hazard that could develop.
Provide advance warning of a plant malfunction that could lead to a health hazard or plant damage. Page: 133 of 200 12/1412009
Provide a warning of any potential inadvertent release of radioactivity to the environment.

The RMS at Farley Nuclear Plant is divided into three subsystems:

Area Radiation Monitoring System (ARMS)
Process and Effluent Radiation Monitoring System (PERMS)
Atmosphere Radiation Monitoring System Previous NRC exam history if any: 072G2.1.27 072 Area Radiation Monitoring System 2.1.27 Knowledge of system purpose and/or function. (CFR: 41.7) RO 3.9 SRO 4.0 Match justification:

Knowledge of what the purposes and functions of the ARMS part of the Radiation monitoring system is required to answer this question. Knowledge is also required of which part of the Radiation monitoring system is comprised of the Area Radiation monitoring system, and what it's purpose(s) and function(s) is(are)-as opposed to the purposes and functions of the PERMS and/or "Atmosphere Radiation Monitoring System". There are only 3 functions/purposes of the RMS system, and all of them apply to the ARMs. To obtain distractors that were plausible but wrong, functions of parts of the RMS were used that were not part of the ARMs subsystem. Chose "shift ventilation to recirc and "secures a liquid release" to ensure the distractors were plausible but definitely wrong (these are accomplished by RMS, but NOT by ARMs). ARMs does provide functions to secure a ventilation system (R-60s) and to secure an airborne release (R-60s). Objective:

1. STATE AND EXPLAIN the purpose and/or function(s) of the Radiation Monitoring System (OPS-40305AOI)

Page: 134 of 200 12/1412009 FNP Units 1 & 2 \ r 3> t;/ ., ,"/\ f{ 7 t:r!_ '}-I\ <L) f'C 3.0 Critical Component Functional Design Requirements 3.1 Area Monitors 3.1.1 3.1.2 3.1.3 3.1.4 3.1.5 3.1.6 3.1.7 3.1.8 3.1.9 3.1.10 3.1.11 3.1.12 Control Room Area Monitor (RE-OOOl) Technical Support Center Area Monitor (RE-OOOIB) Containment Elevation 155'-0" Area Monitor (RE-0002) Radio Chemistry Laboratory Area Monitor (RE-0003) Charging Pump Room Area Monitor (RE-0004) Fuel Storage Pool Area Monitor (RE-0005) Sampling Room Area Monitor (RE-0006) Incore Instrument Area Monitor (RE-0007) Drumming Station Area Monitor (RE-0008) Sample Panel Room Area Monitor (RE-0009) Containment High Range Area Monitor (RE-0027 A, B) Low Level Radwaste Building Area Monitor (RE-0066A, B, C, D, E, F) 3.2 Process Liquid Monitors 3-1 3-1 3-3 3-4 3-6 3-7 3-9 3-10 3-12 3-13 3-14 3-16 Recycle Evaporator Condensate Discharge (RE-OO 16) UNIT 1 -SPARED 3-19 I (J I} Component Cooling Water Pump Suction (RE-0017A, B) 3-23 1 fJ , :\ 3.2.3 Waste Monitor Tank Discharge Radiation Monitor (RE-0018) 3-26 _, ' \1\ 'OJ),.!) Generator Blowdown Sample Radiation Monitor (RE-0019) 3-30 I p' "'\Fontainment Cooler Service Water Outlet (RE-0020A,B) 3-32 \.e/ l; f 3 .. 6 Steam Generator Blowdown (RE-0023A,B) 3-34 ,)' .. 7 Closed Loop Auxiliary Steam System (RE-0026A, B) 3-39 C ) <it ..8 Main Steam Line Nitrogen 16 (N-16) Monitor (RE-0070A, B, C) 3-41 '{'-tv ) 3.3 Particulate, Iodine, and Gas Monitors 3.3.1 3.3.2 3.3.3 3.3.4 3.3.5 3.3.6 3.3.7 3.3.8 3.3.9 3.3.10 3.3.11 Penetration Room Filtration Exhaust Particulate Monitor (RE-OO 1 0) Containment Atmosphere Particulate and Noble Gas Monitor (RE-00I1100I2) Waste Gas Processing System Noble Gas Radiation Monitor (RE-0013) Plant Vent Stack Noble Gas Radiation Monitor (RE-0014) Steam Jet Air Ejector Exhaust Noble Gas Monitor (RE-15, B, C) Plant Vent Stack Particulate and Noble Gas Monitor (RE-002110022) Containment Purge Noble Gas Monitor (RE-0024A, B) Spent Fuel Pool Ventilation Noble Gas Monitor (RE-0025A, B) Steam Jet Air Ejector Exhaust Grab Sampler (RE-0028) Plant Vent Stack Particulate/IodinelNoble Gas Monitor (RE-0029B) Radwaste Area Ventilation Particulate and Noble Gas Monitor, Elevations 100' -0", 83' -0", and 77' -0" (RE-0030A, B) VII J0 3-43 3-43 3-46 3-50 3-53 3-56 3-59 3-64 3-68 3-72 3-73 3-76 Ver.9.0 (a-FNP Units 1 & 2 RADIATION MONITORING SYSTEM A-181015 3.3.12 Radwaste Area Ventilation Particulate Monitor, Elevation 121' -0" (RE-0031) 3-79 3.3.13 Radwaste Area Ventilation Particulate Monitor, Elevation 139' -0" (RE-0032) 3.3.14 Radwaste Area Ventilation Particulate Monitor, Elevation 155'-0" (RE-0033) 3.3.15 3.3.16 3.3.17 3.3.18 3.3.19 3.3.20 3.3.21 Radwaste Area Ventilation Particulate Monitor, Elevation 155'-0" (Access Control Area) (RE-0034) Control Room Ventilation Inlet Noble Gas Monitor (RE-0035A, B) Main Steam Safety Relief and TDAFW Pump Exhaust Noble Gas Monitors (RE-0060A, B, C, D) Plant Vent Stack Particulate/IodinelNoble Gas Grab Sampler (RE-0029A) Containment Post Accident Particulate/IodinelNoble Gas Grab Sampler (RE-0067) Plant Vent Stack Particulate/IodinelNoble Gas Grab Sampler (RE-0068) Containment Purge Exhaust Particulate/IodinelNoble Gas Grab Sampler (RE-0069) 4.0 Structural Design Features 4.1 Shielding Design for Normal Operation 4.2 Shielding Design for Post-Accident Operation 4.3 Isokinetic and Particulate Sample System Design Considerations 4.3.1 4.3.2 4.3.3 Isokinetic Probe Design Considerations Transport Piping Design Considerations USNRC Inspection Criteria 5.0 Noncritical Component Functional Design Requirements 6.0 References 7.0 Index 8.0 Division of Responsibilities V III 3-93 3-96 3-97 3-101 3-104 4-1 4-1 4-1 4-1 4-2 4-2 4-4 5-1 6-1 7-1 8-1 Rev.3 I FNP Units 1 & 2 RADIA nON MONITORING SYSTEM A-181015 The B Train power supply for the RMS system panel Q2H25NGR2504I is 120 VAC distribution panel2K, breaker number 5 (References 6.4.353 and 6.4.354). 3.1.11.7 Shielding Design The detectors shall be located at widely separated locations (References 6.4.020,6.7.005, and 6.7.034). The NUREG 0737 guidance for the location of these detectors states that they should be located to view a large segment of the containment atmosphere. The monitors are calibrated to respond to noble gas energies. The noble gases would be displaced to the upper regions of the containment and the monitors should be situated to obtain the best view of those areas, however, they should not be located such that maintenance would be difficult (i.e., not in the containment dome). Additionally, they should be placed such that radioactive shine from the reactor is minimized. The locations selected for these monitors are on the containment wall approximately 5 feet above floor elevation 155' -0". These locations provide functional shielding by placing the detectors below the 'shine' horizon from the reactor cavity (References 6.4.020 and 6.4.039). 3.1.12 Low Level Radwaste Building Area Monitor 23162.3431A-181015.RM TPNSNo. NSD21RE 0066A through F 3.1.12.1 Basic Function These detectors monitor the low level radwaste buildings to alert operators to an increase in radiation levels as required by GDC 63 (Reference 6.7.084). 3.1.12.2 Functional Requirements The monitors shall provide continuous indication over a range of 0.1 to 10 4 millirads per hour (mRlhr). (Reference 6.4.135). 3.1.12.3 I&C Requirements 3.1.12.3.1 The monitors shall provide a flat (+/-15 percent) response for gamma energies between 40 ke V and 1.25 MeV (Reference 6.7.052). 3-18 Rev. 0 FNP Units 1 & 2 RADIATION MONITORING SYSTEM A-181015 3.1.12.3.2 The monitor ratemeter shall provide a minimum of two single pole double throw alarm relays for external use. Each relay shall be fully adjustable over the entire indicated range of the monitor (Reference 6.7.052). {etX' Upon detection of gross activit in excess of the trip setpoint, the monitors shal stop he ventilation system fans (Reference . .12 . I If.-UJ' 3.1.12.3.4 The setpoint for this monitor is based on being high enough to prevent spurious alarms, but low enough to alert the personnel in the low level radioactive storage building to an increase in radiation levels (Reference 6.7.080). 3.1.12.4 Interface Requirements The power supply for the Radiation Level Indicator Panel NSD21 G502 is 120/208 V AC distribution cabinet 1 TT, breaker number 14 (Reference 6.4.130). 3.2 PROCESS LIQUID MONITORS 3.2.1 Recycle Evaporator Condensate Discharge 23162.343IA-181015.RM TPNS No. ND 11 REOO 16 Unit 1 detector ND 11 RE 0016 has been spared and is abandoned in place. Control and alarm functions of the detector have been disconnected and disabled. The detector remains physically installed in the discharge line from the recycle evaporator. 3.2.1.1 Basic Function This detector monitors the effluent from the recycle evaporator and directs it to either the Reactor Makeup Water (RMW) System or the Recycle Holdup tanks to minimize contamination of the RMW system (References 6.4.358 and 6.7.080). 3.2.1.2 Functional Requirements 3.2.1.2.1 An in-line liquid monitor shall be provided to directly monitor the process medium. The use of this type of monitor provides the fastest response time and easiest decontamination (References 6.4.358 and 6.7.080). 3-19 Ver.9.0 FNP Units 1 & 2 RADIATION MONITORING SYSTEM A-181015 3.2.2.5.2 The A train DC control power supply for the main control board panel NH11NGMCB 2500A-AB is 125 VDC distribution panellA, breaker number 15 (Reference 6.4.105). The B train DC control power supply for the main control board panel NHIINGMCB 2500A-AB is 125 VDC distribution panel ID, breaker number 12 (Reference 6.4.115). 3.2.3 Waste Monitor Tank Discharge Radiation Monitor 23162.343IA-181015.RM TPNS No. NDllRE 0018 3.2.3.1 Basic Function This detector monitors the radioactive waste processing system common liquid effluent line while waste is being discharged to the environment and initiates automatic isolation of the discharge if setpoints are exceeded to comply with GDC 60 and GDC 64 (References 6.4.038 and 6.7.084). 3.2.3.2 Functional Requirements 3.2.3.2.1 An in-line liquid monitor shall be provided to directly monitor the process medium. The use of this type of monitor provides the fastest response time and easiest decontamination (References 6.4.038, and 6.7.080). 3.2.3.2.2 This monitor shall alarm and isolate the effluent discharge prior to exceeding the limits of ten times the concentrations stated in 10 CFR Part 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in unrestricted areas will result in exposures within (1) the Section ILA design objectives of Appendix I, 10 CFR 50, for a member of the public, and (2) the limits of 10 CFR 20.1301 for the population. The limiting concentration of dissolved and entrained noble gases is 1 x 10-4 /lCi/ml (References 6.7.078 and 6.7.081). Setpoints are based on ensuring the discharge limits presented in Section 2.1.2 of the ODCM are not exceeded. 3-26 Rev. 0 FNP Units 1 & 2 RADIA TION MONITORING SYSTEM A-181015 inputting to a recorder (References 6.4.109, 6.4.318, 6.7.079, and Open Item Observation RMS-FSD-014). 3.3.15.3.3 The monitor ratemeter shall provide a minimum of two single pole double throw alarm relays for external use. Each relay shall be fully adjustable over the entire indicated range of the monitor. One relay shall actuate on a high signal to actuate the local annunciator. In addition, each function shall actuate an indicating light on the monitor front panel defining the alarm condition. A power on/off switch shall be provided for the ratemeter (Reference 6.4.318). 3.3.15.3.4 The sampler control panel is mounted in the monitor skid and shall provide switches for detector power on/off, pump stop/start, and an alarm light indicating low flow conditions in the monitor. A local control panel shall be provided at the monitor skid that contains switches for pump operation and indicating lights providing alarm and fail status (Reference 6.4.318). 3.3.15.3.5 The setpoints for this monitor are calculated to ensure that the limits specified by 10 CFR 20, Appendix B, Note 2, Table 1, Column 3 are not exceeded (References 6.7.062 and 6.7.083). 3.3.15.4 Interface Requirements The 120 VAC power supply for RMS panel NSDIIRE 0034 is 2081120 VAC control power panel T, breaker number 12 (Reference 6.4.108). The 208 V AC pump power supply for RMS panel NSD 11 RE 0034 is 2081120 VAC control power panellDD, breaker number 14 (Reference 6.4.108). 3.3.16 Control Room Ventilation Inlet Noble Gas Monitor 23162.3431A-181015.RM TPNS Nos. QSDllRE 0035A, B 3.3.16.1 Basic Function These detectors provide redundant, safety related, monitoring of the outside air entering the control room through the computer room air intake and to initiate automatic isolation of the control room ventilation 3-89 Rev. 0 FNP Units 1 & 2 RADIATION MONITORING SYSTEM A-18101S system dampers (HV -3622, 3624, 3626, and 3628 for train A and HV-3623, 362S, 3627, and 3629 for train B) if setpoints are exceeded. This air inlet is common to the main control room and is isolated to comply with the criteria in GDC 19. In addition, the T C ventilation s stem is ill!tomaticall shifted to recirc mode (References 6.4.377, 6.4.378,6.4.379,6.4.380, 6.4.38S, 6.7.001, and 6.7.084). For a complete description of the operation of the Control Room Ventilation Systems, see the Control Room Ventilation System FSD, Drawing No. A-181006. For a complete description of the operation of the Technical Support Center Ventilation Systems, see the Auxiliary Building Ventilation System FSD, Drawing No. A-181016. 3.3.16.2 Functional Requirements 3.3.16.2.1 The monitor shall consist of an off-line gas monitor located in an area of low background radiation (References 6.4.290, 6.4.349, and 6.7.080). 3.3.16.2.2 The monitor shall provide a pumping system with a sample flow rate of 8.S scfm. The sample shall pass through a pre-filter to remove particulates larger than S microns prior to entering the volume chamber for counting (Reference 6.4.317). The flow rate of 8.S scfm is the manufacturer's standard design. In the present configuration, this flow rate provides a monitor response and control room intake isolation time of approximately 7.4 seconds. The fuel handling accident analysis inside containment credits RE-03SA and B for isolation of the control room due to high radiation within 60 seconds following accident initiation (References 6.3.004 and 6.3.006). 3.3.16.2.3 This monitor's function is to detect radioactive gases, however, certain elements of isokinetic design have been maintained in the transport system design. The tubing leading up to the monitor inlet contains radius bends in excess of five times the tube diameter. As described in FSD Section 4.3, this is to plateout of particulates in the sample lines. A sample probe is included on the inlet tube at the duct interface. The probe is not an isokinetic design. The probe provides a beveled side which is placed to face into the upstream air flow. This prevents a sheer across the face of the probe tip which could create a vacuum and decrease the sample flow rate. The system is operated in an ani so kinetic mode with sample velocity 3-90 Rev. 3

51. 073K3.01 001/N8N/RO/C/A 3.6/4.2/N/N/2/HBF/Y Given the following plant conditions:

Unit 1 is in Mode 5.
Spent Fuel is being moved in preparation for a refueling outage.
R-25A, SFP VENT, radiation monitor loses instrument power. Which one of the following describes:

1) the Train(s) of PRF RECIRC and EXH fans that automatically start, and 2) whether or not manual action is required to OPEN HV-3538A, SFP to 1 A PRF SUPPLY DMPR? A. 1) Only A train starts; 2) Manual action is required.

B. 1) Only A train starts; 2) Manual action is NOT required. C. 1) Both trains start; 2) Manual action is required. 1) Both trains start; 2) Manual action is NOT required. Page: 135 of 277 12/14/2009 A -Incorrect. Plausible: A -Incorrect. C -Incorrect. D -Correct. 1 A PRF system will be started directly due to the R-25A rad monitor failure, but as a result of the SFPR Differential pressure, the 1 B PRF train is expected to start also. HV3538A should already be open for the given plant conditions. Action is not required to align it. 1) 1 A train PRF is directly started from R-25A; a separate start signal is provided to the 1 B Train PRF system. 2) HV3538A & B are required to be verified open following an autostart of the PRF system per P&L 3.3 of SOP-58.D. They do not automatically open. Applicant may be aware that the dampers don't automatically operate with an alarm on R-25A or on low dip SFP ventilation but be confused on the normal position. See A for discussion and plausibility. See A for discussion and plausibility. See A for discussion.

REFERENCES:

SOP-45.0, ver 36.0, P&L 3.5 "The radiation monitors fail to a "High Radiation" conditions on a loss of instrument and/or control power that will result in actuation of associated automatic functions. [ ... ]" ARP-1.6, vers 64.0, FH1 and FH5; R-25A & B automatically trips the SFP Supply AHU, both EXH Fans and closes the supply and exhaust dampers. And starts the associated train PRF. Additionally, the unaffected train penetration room filtration system will start due to Low .6.P in the spent fuel pool room. SOP-58.0, ver 70.0, Step 3.9: "PRF system auto start form R-25A or R-25B requires operator action to verify open SFP TO 1A PRF SUPPLY DMPR, [ ... ] or SFP TO 1 B PRF SUPPLY DMPR,[ ... ]." TSR 3.7.12 (REQUIRED during SFP movement in the SFPR) VERIFY two PRF trains aligned to the SFPR. Page: 136 of 277 12/14/2009 Previous NRC exam history if any: (MODIFIED? NEW?) MODIFIED FROM AUX BLDG VT-40304B07 015 ----2006 NRC MODIFIED FROM AUX BLDG VT-62107B01 004 ----none MODIFIED FROM RMS-40305A07 003 ----2001 NRC 073K3.01 073 Process Ra::liation Monitoring System K3 KnoNIedge d the effect that a IO$Q" malfundion d the PRM will have on the follCMfing: (CFR: 41.7 /45.6) K3.01 Radiooctive effluent raEB5ElS ...................................... 3.64.2 Match justification: -R-25A & B require PROCESS flow (SFP system flow) to be operable therefore they are considered PROCESS RADIATION MONITORs. -SFP HVAC effluent is discharged to the Plant Vent stack via the Aux Bldg Main Exh fans and the plenum. PRF discharges directly to the Plant Vent Stack, therefore these systems can be considered Radioactive Effluent Release paths. -R-25A failure is indicated in the stem which impacts (affects) that radioactive effluent release. Somewhat related to a SirnulatQ" Scenario (#2) failure on this exam, but the Sc61ario has a R-25A hi alarm (i nstea:l of an i nstrUmalt power fai lure), with auto SFP V 61ti I ati on i 001 ati on defeated, whi ch wi II causa onl y one trai n of PRF to arto start i ni ti all Y (the other won't auto start unti I the SFP venti I ati on is manual I y 93CUrEd). Thi s question, tests en insrurnent poNer failure with no failure of the SFP v61tilation i001 ation which 93CUres SFP v61ti I ati on end the low dip a:;ross the SFP venti I ati on fans starts both PRF trai ns i ni ti all y. Thi s questi on tests the knowledge of how afailure of one R-25 affoc1:sthe PRF &ystern with no failure of SFPv61tilation &ystern to auto 93Cure. Objective: OPS-40304B02; Relate and Identify the operational characteristics including design features, capacities and protective interlocks for the components associated with the Auxiliary Building Ventilation Systems [ ... ] Page: 137 d 277 12/14/2009 Question # 51 KIA 073K3.01 REFERENCE Docs 04/07/08 13:33:43 FNP-I-S0P-45.0 2.0 Initial Conditions 2.1 The electrical distribution system is energized and aligned for normal operation per FNP-I-S0P-36.0, PLANT ELECTRICAL DISTRIBUTION LINE-UP, with exceptions noted. 2.2 120V AC electrical distribution system is energized and aligned for operation per FNP-I-SOP-36.4, 120V A.C. DISTRIBUTION SYSTEMS. 2.3 Power fuses are installed in all radiation monitoring system instrument drawers. 2.4 The radiation monitoring system is aligned per System Checklist FNP-l-SOP-45.0A. 2.5 Containment radiation monitors R-Il and R-12 are aligned for operation per FNP-I-SOP-12.2, CONTAINMENT PURGE AND PRE-ACCESS FILTRATION SYSTEM (applies only to R-il and R-12). 2.6 Have I&C verify proper monitor alignment per FNP-I-IMP-227.47, VALVE ALIGNMENT FOR OPERATION OF UNIT 1 REOOI 1112 AND RE0021122. 3.0 Precautions and Limitations 3.1 Due to slow filter paper speed of the APD a five hour time period is required for the detector indication to reach equilibrium value after changing filter paper or filter paper speed. 3.2 Alarms on Radiation Monitoring System Panel must be acknowledged to provide main control board annunciator reflash capability. 3.3 A common annunciator on the main control board is actuated on high radiation from any channel. Individual drawers shall be checked to determine the alarming channel(s). 3.4 A common annunciator on the main control board is actuated when any channel is in the test mode. 3.5 The radiation monitors fail to a "High Radiation" condition on loss of instrument and/or control power that will result in actuation of associated automatic functions. Prior to removing a channel from service, ensure any automatic functions are either disabled or acceptable with respect to the affect (including reportability consideration) on associated system operation. (Refer to Tables A, Band C for monitors with automatic functions.) Version 34.0 04/03/09 13 :21 :48 FNP-I-ARP-l.6 LOCATION FH 1 RADIA TION MONITOR REFERENCE TABLE (cont) RE LOCATION TYPE DETECTOR FUNCTION ACTIONS R-21 Plant Vent Stack (AB 155') APD Scint. Perform Step (Victoreen) 4.21 R-22 Plant Vent Stack Gas G-M(W Perform Step ODCM (AB 155') ) 4.22 R-23A SG Blowdown Surge Tank Liquid Scint. (W) Closes Perform Step Inlet (AB 130') FCV-1152 4.23 R-23B SG Blowdown Surge Tank Liquid Scint. (W) Closes Perform Step ODCM Discharge (AB 130') RCV-23B 4.23 R-24A* Containment Purge (AB Gas Scint. Closes No input to 155') containment this alarm purge supply & exhaust dampers 2866C & 2867C and 3198A & 0 R-24B* Containment Purge Gas Scint. Closes valves: No input to (AB 155') (Victoreen) 28660 &28670, this alarm 3196,3197, 3198B & C R-25AI Spent Fuel Pool Ventilation Gas Scint. Trip fuel bldg No input to R-25B* (AB 184') (Victoreen) supply and this alarm exhaust fans; closes SFP HVAC supply and exhaust dampers; starts associated trains of penetration room filtration. R-26A Recycle Evap. Condo Liquid Scint. (W) Perform Step Recovery Unit (AB 100') 4.24 R-26B Waste Evap. Condo Liquid Scint. (W) Perform Step Recovery Unit (AB 100') 4.24 R-27A* Containment (High Range) Area Ion Chamber No input to (Victoreen) this alarm Technical Specification related Page 6 of 12 Version 59.0 04103109 13 :21 :48 SETPOINT: Variable, as per FNP-I-RCP-252 ORIGIN: Radiation Monitor Cabinet Channels R-25A or R-25B, Spent Fuel Pool Vent PROBABLE CAUSE H5 FNP-I-ARP-l.6 LOCATION FH5 SFP AREA RE25 A OR B HIRAD I. High Radiation Level in the discharged air from the Spent Fuel Pool Area Ventilation Fans. 2. The radiation monitors fail to a "High Radiation" condition on loss of instrument and/or control power that will result in actuation of associated automatic functions. AUTOMA TIC ACTION NOTE: The unaffected train penetration room filtration system may also start, due to low in the spent fuel pool. Trips the Fuel Handling Area Supply and Exhaust Fans, closes the Fuel Handling Area Supply and Exhaust Dampers AND starts the Penetration Room IA OR IB Filtration Units. OPERA TOR ACTION 1. Determine which radiation monitor indicates high activity.

2. Verify that the required automatic actions listed above have occurred.

IF any automatic actions have not occurred, THEN go to FNP-I-SOP-58.0. (The section for Fuel Handling Area Heating and Ventilation Operation for guidance)

3. Announce receipt of the alarm and the affected area on the public address system. 4. Have all personnel evacuate the affected area. 5. Implement FNP-0-ElP-9, EMERGENCY CLASSIFICATION AND ACTIONS. 6. Determine the validity of the high activity indication as follows: 6.1 Verify that the instrument is aligned for normal operation and is functioning properly.

6.2 Sample or survey the affected system or area as required.

7. Determine the source or cause of the high activity and correct or isolate as required.

8. DO NOT allow personnel to enter the affected area without the approval of the Health Physics Department.

9. IF high activity indication is due to instrument failure, THEN refer to Technical Specifications, section 3.3.8. {CMT 0008659} applies to entire annunciator.}

References:

A-I77100, Sh. 310; U-258400; 0-181658; 0-181671; 0-177394 Sh. 1&2; FSAR, Section 11.4; 0-175045 Page 1 of 1 Version 59.0 11125/0807:46:06 FNP-I-SOP-58.0 3.0 Precautions and Limitations 3.1 The 600V Load Center Dampers are controlled by outside air temperature (NIV47TSHL3633) and this feature should not be defeated:

Dampers close when temperature:::::

60° F o Dampers open when temperature

s 56° F 3.2 Placing an engineered safety features pump in local control places its respective pump room cooler in automatic only. 3.3 PRF System auto start from R-25A or R-25B requires operator action to verify open SFP TO lA PRF SUPPLY DMPR, QIV48HV3538A or SFP TO IB PRF SUPPLY DMPR, QIV48HV3538B.

3.4 Auxiliary Building Battery Charger Room Coolers are no longer considered attendant equipment. TS 3.7.19, ESF Room Coolers, is now applicable, and Auxiliary Building Battery Charger Room Coolers are considered support equipment. A battery charger room cooler can only support one operating charger. For this situation, operating means in service, load testing, in general when producing heat. 3.5 When Service Water is unavailable to these room coolers, their associated fan motor must remain tagged out to prevent the fan from running and adding a heat load to the associated room. o CCW Pump Room

Battery Charger Room
IA/IB MCC Room 3.6 600V Load Center ID & IE Room Coolers are no longer considered attendant equipment.

TS 3.7.19, ESF Room Coolers, is now applicable, and 600V Load Center I D & 1 E Room Coolers are considered support equipment. Version 67.0 11125/08 07:46:06 FNP-I-SOP-58.0 4.8 Fuel Handling Area Heating and Ventilation Operation. CAUTION: One train of PRF must be in operation per step 4.8.2.1 if SFP HV AC is not in service. (SFP HV AC shuts down or the SFP EXH FAN SUCT DMPR Q1 V 48HV3990A(B) for the running fan is found Closed.) NOTE: BOTH trains ofPRF should be started and Spent Fuel Pool Ventilation secured prior to sipping known leaking fuel assemblies in the Spent Fuel Pool to prevent radioactive gas release causing R-25A or B alarm and subsequent PRF auto start. (CMT 10518) 4.8.1 To place the fuel handling area heating and ventilation in service, perform the following: CAUTION: Both Auxiliary Bnilding Main Exhaust fans must be in operation if Containment Purge and SFP Exhaust fans are running sim ultaneously. NOTE: Exhaust fan will shutdown after 20 seconds if dampers are not open in step 4.8.1.2. 4.8.l.1 Start one of the following fans.

lA SFP EXH FAN, NIV48MOOIA
IB SFP EXH FAN, NIV48MOOIB 4.8.1.2 Open the following:
SFP EXH FAN SUCT DMPR, QIV48HV3990A
SFP EXH FAN SUCT DMPR, QIV48HV3990B
SFP AHU DISCH TO SFP QIV48HV3991A
SFP AHU DISCH TO SFP Q I V48HV3991 B Version 67.0 11125/08 07:46:06 4.8.1.3 4.8.1.4 4.8.1.5 FNP-I-S0P-58.0 Start SFP AHU SUPP FAN, NIV48M002.

Verify open the following:

SFP TO PRF FLTR UNIT QIV48HV3538A
SFP TO PRF FLTR UNIT QIV48HV3538B IF required, THEN secure any running train(s) ofPRF per FNP-I-S0P-60.0, PENETRATION ROOM FILTRATION SYSTEM. NOTE: BOTH trains ofPRF should be started and Spent Fuel Pool Ventilation secured prior to sipping known leaking fuel assemblies in the Spent Fuel Pool to prevent radioactive gas release from causing the R-25A or R-25B alarm and subsequent PRF auto start. (CMT 10518) 4.8.2 To remove fuel handling area heating and ventilation from service, perform the following:

4.8.2.1 IF 'A' train PRF is to be run AND aligned to SFP Room, THEN perform the following:

1. Verify requirements of Step 3.16 and 3.19 are met. NOTE: Technical Specification SR 3.7.12.1 requires both trains ofPRF aligned to the spent fuel pool room during movement of irradiated fuel assemblies in the spent fuel pool room. The following step is N/A during fuel movement.

2. Close SFP TO IB PRF SUPPLY DMPR, QIV48HV3538B.

3. Verify open SFP TO IA PRF SUPPLY DMPR, QIV48HV3538A

4. Start IA PRF RECIRC FAN, QIEI5M002A.

5. Start lA PRF EXH FAN, QIE15MOOIA

6. IF Ql V48HV3538B NOT closed THEN, start 'B' train PRF as follows: a. Start IB PRF RECIRC FAN, QIE15M002B

b. Start IB PRF EXH FAN, QIE15MOOIB Version 67.0 PRF 3.7.12 3.7 PLANT SYSTEMS 3.7.12 Penetration Room Filtration (PRF) System LCO 3.7.12 Two PRF trains shall be OPERABLE.

NOT E ---------------------------------------------- The PRF and Spent Fuel Pool Room (SFPR) boundaries may be opened intermittently under administrative control. APPLICABILITY: MODES 1, 2, 3, and 4 for post LOCA mode of operation, During movement of irradiated fuel assemblies in the SFPR for the fuel handling accident mode of operation. ACTIONS CONDITION A. One PRF train inoperable. B. Two PRF trains inoperable in MODE 1, 2, 3, or 4 due to inoperable PRF boundary. C. Required Action and associated Completion Time of Condition A or B not met in MODE 1, 2, 3, or 4. OR Two PRF trains inoperable in MODE 1, 2, 3, or 4 for reasons other than Condition B. D. Required Action and associated Completion Time of Condition A not met during movement of irradiated fuel assemblies in the SFPR. Farley Units 1 and 2 A.1 B.1 C.1 A.ND C.2 D.1 OR D.2 REQUIRED ACTION COMPLETION TIME Restore PRF train to 7 days OPERABLE status. Restore PRF boundary to 24 hours OPERABLE status. Be in MODE 3. 6 hours Be in MODE 5. 36 hours Place OPERABLE PRF Immediately train in operation. Suspend movement of Immediately irradiated fuel assemblies in the SFPR. 3.7.12-1 Amendment No. 161 (Unit 1) Amendment No. 154 (Unit 2) ACTIONS CONDITION REQUIRED ACTION PRF 3.7.12 COMPLETION TIME E. Two PRF trains inoperable E.1 during movement of Suspend movement of Immediately irradiated fuel irradiated fuel assemblies assemblies in the SFPR. in the SFPR. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.12.1 ----------------------------NOTE---------------------------------- Only required to be performed during movement of irradiated fuel assemblies in the SFPR. ---------------------------------------------------------------------- Verify two PRF trains aligned to the SFPR. 24 hours SR 3.7.12.2 Operate each PRF train for;:: 15 minutes in the 31 days applicable mode of operation (post LOCA and/or refueling accident). SR 3.7.12.3 Perform required PRF filter testing in accordance In accordance with with the Ventilation Filter Testing Program (VFTP). the VFTP SR 3.7.12.4 Verify each PRF train actuates and the normal spent 18 months fuel pool room ventilation system isolates on an actual or simulated actuation signal. SR 3.7.12.5 Verify one PRF train can maintain a pressure 18 months on a ::;; -0.125 inches water gauge with respect to adjacent STAGGERED areas during the post LOCA mode of operation at a TEST BASIS flow rate::;; 5500 cfm. SR 3.7.12.6 Verify one PRF train can maintain a slightly negative 18 months on a pressure with respect to adjacent areas during the STAGGERED fuel handling accident mode of operation at a flow TEST BASIS rate::;; 5500 cfm. Farley Units 1 and 2 3.7.12-2 Amendment No. 161 (Unit 1) Amendment No. 154 (Unit 2)

1. AUX BLDG VT-40304B07 015 Given the following plant conditions:

Unit 1 is Mode 5.
Containment purge is in operation.
STP-50, Radiation Monitor Monthly Source Check, is being performed.
R-25A, SFP VENT, radiation monitor starts increasing.

Which one of the following describes the system response if R-25A exceeds the alarm setpoint? A. The automatic actions of R-25A will be blocked while performing this STP. B. The 1 A fuel handling area supply and exhaust fans trip and the 1 A fuel handling area supply and exhaust dampers close. The 1 B fuel handling area supply and exhaust fans start. C"!' The fuel handling area supply and exhaust fans trip, the fuel handling area supply and exhaust dampers close and the penetration room 1A and 1 B filtration units start. D. The fuel handling area supply and exhaust fans trip, the fuel handling area supply and exhaust dampers close and the containment purge supply and exhaust valves close. Feedback FH5 SFP AREA RE25 A OR B HI RAD A. Incorrect-If the alarm circuit actuates during the STP the automatic functions provided by the instrument will occur. B. Incorrect-The A fuel handling area fans will trip, the fuel handling area dampers will close, but penetration room filtration units will start, not the other train of SFP HVAC. C. Correct -These are the correct automatic actions that will occur when R-25A alarms. FH5 SFP AREA RE25 A OR B HI RAD Automatic action: Trips the Fuel Handling Area Supply and Exhaust Fans, closes the Fuel Handling Area Supply and Exhaust Dampers AND starts the Penetration Room1A OR 1 B Filtration Units. When R25A alarms, 1A PRF starts. When the Fuel handling area supply and exhaust dampers close, the other train of PRF starts. Lesson plan auto starts: During normal operation, the PRF system is aligned to automatically process the exhaust air from the spent fuel pool upon receipt of an actuation signal initiated by either (1) high radiation or (2) low flow in the spent fuel pool exhaust system. D. Incorrect-The fuel handling area fans will trip, the fuel handling area dampers will close, but the containment purge valves will not close. Page: lof6 9/1712009 Notes 2006 NRC exam KIA: Area Radiation Monitoring (ARM) System -Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ARM system controls including: Radiation levels. Page: 2 of6 911712009

2. AUX BLDG VT-62107BOI 004 Given the following plant conditions:

Unit 1 is Mode 6 and refueling operations are in progress.
Containment purge is in operation.

During a Channel Operational Test (COT) on R-25B, SFP VENT, the trip setpoint was found to be out of tolerance high and this was reported to the Shift Manager. After this report was made R-25A, SFP VENT, failed high and the red alarm light is illuminated. Which one of the following describes the system response following the failure and the actions that fully meet the required actions lAW Tech Specs? The Spent Fuel handling area supply and exhaust fans (A) The Spent Fuel handling area supply and exhaust dampers (B) Immediately (C) A. (A) trip (B) remain open (C) ensure BOTH PRF trains are operating B. (A) remain running (B) close (C) secure fuel handling in the SFP room (A) trip (B) close (C) ensure BOTH PRF trains are operating D. (A) remain running (B) remain open (C) secure fuel handling in the SFP room Feedback TS 3.3.8 immediate ACTION completion times SRO level due to 55.43 b 2 A. Incorrect-The fuel handling area supply and exhaust fans trip, the fuel handling area supply and exhaust dampers close, and 1A and 1 B PRF units start due to the high alarm on R-25A. This is due to the following, when R25A alarms, 1A PRF starts. When the Fuel handling area supply and exhaust dampers close, the other train of PRF starts. B. Incorrect-Both trains of PRF start. The proper TS action is to ensure both trains of Page: 3 of6 9/1712009 PRF start. If this can not be done then condition C will have the fuel movement stopped. C. Correct -These are the correct automatic actions that will occur when R-25A alarms. FH5, SFP AREA RE25 A OR B HI RAD Automatic action: Trips the Fuel Handling Area Supply and Exhaust Fans, closes the Fuel Handling Area Supply and Exhaust Dampers AND starts the Penetration Room 1A OR 1 B Filtration Units. This is also the TS action for both rad monitors OOC. The candidate has to know that a failed COT will render the rad monitor inoperable and the one failed high will cause the automatic actions. The response to ensure Both trains of PRF are running is lAW TS 3.3.8. When R25A alarms, 1A PRF starts. When the Fuel handling area supply and exhaust dampers close, the other train of PRF starts. Normally the supply fan and one exhaust fan, A or B, are in operation. Lesson plan auto starts: During normal operation, the PRF system is aligned to automatically process the exhaust air from the spent fuel pool upon receipt of an actuation signal initiated by either (1) high radiation or (2) low flow in the spent fuel pool exhaust system. D. Incorrect-The fuel handling area fans will trip, the fuel handling area dampers will close, and both the Penetration Room 1A and 1 B Filtration Units will be operating. This is the incorrect TS action. Proper use of Tech specs would require first attempting condition B before attempting C. Also the automatic actions have placed both PRF trains in service, so Condition B is met. There would be no reason to enter condition C. Bases 3. Spent Fuel Pool Room Radiation The LCO specifies two required Gaseous Radiation Monitor channels to ensure that the radiation monitoring instrumentation necessary to initiate the PRF remains OPERABLE. Each monitor will initiate the associated train of PRF and isolate the normal Spent Fuel Pool Room ventilation. For sampling systems, channel OPERABILITY involves more than OPERABILITY of channel electronics. OPERABILITY requires correct valve lineups, sample pump operation, and detector OPERABILITY. The most common cause of channel inoperability is outright failure or drift of the bistable or process module sufficient to exceed the tolerance allowed by unit specific calibration procedures. Typically, the drift is found to be small and results in a delay of actuation rather than a total loss of function. This determination is generally made during the performance of a COT, when the process instrumentation is set up for adjustment to bring it within specification. If the Trip Setpoint is less conservative than the tolerance specified by the calibration procedure, the channel must be declared inoperable immediately and the appropriate Condition entered. Page: 40f6 9/1712009 Notes KIA 072 Area Radiation Monitoring A2.02 Ability to (a) predict the impacts of the following malfunctions or operations on the ARM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Detector failure. KIA MATCH ANALYSIS R-25A, is a ventilation monitor for the area of the SFP and is inoperable due to a detector failure. Tech Specs are impacted by this failure and must be complied with to correct, control, or mitigate the consequences. The question is SRO-only level because it requires basis knowledge to arrive at the answer, in which the candidate needs to know what the affects are of a failed COT which is defined in TS bases. Page: 50f6 911712009

3. RMS-40305A07 003 Annunciator FH5, SFP AREA RE25 A OR B HI RAD, is in alarm on Unit 1. It has been determined that Spent Fuel Pool (SFP) Exhaust Flow Gas monitors R-25A and R-25B indicate high activity.

Which ONE of the following describes the automatic action(s) that occur as a result of this alarm? The SFP supply and exhaust fans (A) ,the SFP HVAC supply and exhaust dampers close, and (B) A. (A) remain running (B) penetration room filtration units do NOT start B. (A) remain running (B) BOTH penetration room filtration units 1A and 1 B start C. (A) trip (B) the penetration room filtration units do NOT start D'!' (A) trip (B) BOTH penetration room filtration units 1A and 1 B start __________ Feedback A -Incorrect, These fans trip and will not remain running as ctmt purge and minipurge fans do. This is plausible since on an alarm from R-24A and B the mini purge fans and ctmt purge fans will remain running and the valves will close and PRF does not start. B -Incorrect, These fans trip and with both alarms R-25 A & B in alarm, both filtration units will start. C -Incorrect, see B above o -Correct-These fans trip and BOTH trains of PRF starts Notes 2001 nrc exam Page: 6of6 9/1712009

52. 074EA 1.28 005lNBN/RO/C/A 3.7/3.9/N/N/3IHBFN The following plant conditions exist on Unit 1:

FRP-C.2, Response to Degraded Core Cooling, is in progress.
All attempts to establish HHSI flow were unsuccessful.
SG depressurization to 100 psig has been completed.
CETCs are 725°F and steady.
All ECCS disconnects have been closed.
B Train SI could NOT be reset, MLB 1 11-1 remains lit.
The OATC has placed all SI Accumulator Discharge Isolation valves in the closed position.
The following are the indications available on the MCB: II WH-LiT I II WH-.DRK .* 1 IG\t-.DRKII R-LIT I IGRoDRKl1 R*LlT I 18ACCUM DISCH ISO Q1 E21 MOV88088 1CACCUM DISCH ISO Q1 E21 MOV8808C LEGEND: WH = WHITE R =RED GR = GREEN LIT = ILLUM I NA TED DRK= NOT ILLUMINATED Neither of the valves are responding to MCB switch manipulation.

Which one of the following describes the reason? MOV-8808B MOV-8808C A. Supply breaker has tripped open. B Train SI is NOT reset. B Train SI is NOT reset. C. RCS pressure is too high. Supply breaker has tripped open. Supply breaker has tripped open. D. Supply breaker has tripped open. RCS pressure is too high. A -Incorrect.

1) The white light LIT indicates control power availability to the MOV, and normally would be sufficient to allow for operation; MOV-8808B can not be closed until after the SI signal is also reset, the SI signal "locks out" a close Signal to these valves. (K603 relay) 2) MOV-8808C is an A train component, B train SI does not block its closure via the (K603) contact. Furthermore, the Breaker is tripped as indicated by the white light. Plausible:

1) Many switches on the MCB are equipped with an AMBER light above the position indication to indicate a tripped supply breaker, this light Page: 138 c:l2n 1211412009 B-Correct. NU I lit Indicates "proper" operation.

2) Both trains of SSPS recieve information from single indications within the field --this concept could be reversed thinking that both trains of SSPS provide input to all 3 components to allow closure (ex: as both trains of 125VDC are required for MFIV opening or SD operation).

Also, since plant is equipped with 3 Accumulators train alignment is necessary to differentiate the impact from SI not being reset. 1) MOV-8808B can not be closed until after the B Train SI signal is also reset, the SI signal "locks out" a close signal to these valves. (K603 relay). 2) MOV-8808C has experienced a loss of control power the motor power contacts as indicated by the loss of the White light. C -Incorrect.

1) Following the SG depressurization, RCS pressure is very likely well below the 2000 psig auto-open signal. Furthermore, this signal does not prevent closure, it would only re-open the valve if were closed psig with an SI signal present. Plausible:

If pressure was >2000 psig and SI was not reset, then upon closure, an auto-open signal would be present; K628 relay (>P-11) and K603 relay (SI) would re-open the valve. If incorrectly applied CETs to RCS temp, and a saturated conditoin assumed then pressure in the RCS could be presumed to be very high--not considering the SG cooldown to 100 psig. 2) this part is correct see B#2. D -Incorrect.

1) See A. 2) equivalent to discussion in C for the wrong train component.

Page: 139 d 277 12/14/2009 Page: 140 of 2n 12/14/2009 Previous NRC exam history if any: NONE 074EA1.28 074 Inadequate Core Cooling EA1 Ability to operate and monitor the following as they apply to a Inadequate Core Cooling: (CFR 41.7 / 45.5 /45.6) EA 1.28 Core flood tank isolation valve controls and indicators ................. 3.7* 3.9* Match justification:

the ability to operate: knowledge of the system interlocks and controls is required to ensure suscessful operation of a component. (the operation is already completed as an initial condition)
ability to monitor: Given the indications, the examinee must evaluate response to validate completion success.
AS they apply to Inadequate core cooling: actions required by FRP-C.2 in a "degraded" core cooling situation (which is a condition resulting from inadequate cooling) --equivical to C.1 actions Objective:

OPS-52533C07 Analyze plant conditions and determine the successful completion of any step in FRP-C.2 Page: 141 cl2n 12/14/2009 Question # 52 KIA 074EA1.28 REFERENCE Docs

FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response II 8 Check core cooling. 8.1 Check REACTOR VESSEL LEVEL indication -GREATER THAN 0% UPPER PLENUM. 8.2 exit TiCs -LESS 8.3 Go to procedure effect. 9 Check SI accumulator valve status. 9.1 Check power to discharge valves -AVAILABLE. 1A(lB,lC) ACCUM DISCH ISO [] Q1E21MOV8808A [] Q1E21MOV8808B [] Q1E21MOV8808C 9.2 Check discharge valves -OPEN. 1A(lB,lC) ACCUM DISCH ISO [] Q1E21MOV8808A [] Q1E21MOV8808B [] Q1E21MOV8808C 10 Monitor CST level 10.1 [CA] Check CST I than 5.3 ft. CST LVL [] LI 4132A [] LI 4132B 10.2 Align makeup to the CST from water treatment plant OR demin water system using FNP-1-S0P-5.0, DEMINERALIZED MAKEUP WATER SYSTEM, as necessary. _Page Completed Page 11 of 22 Response NOT Obtained 8.1 IF SI established, THEN return to step 2, IF NOT, proceed to step 9. 8.2 IF core exit TiCs falling, THEN return to step 2, IF NOT, proceed to step 9. 9.1 Close accumulator dischar valve disconnects using ATTACHMENT

1. 9.2 IF accumulators have NOT discharged, THEN open discharge 1A (lB, 1C) ACCUM DISCH ISO [] Q1E21MOV8808A

[] Q1E21MOV8808B [] Q1E21MOV8808C 10.1 Align AFW pumps suction to SW using FNP 1 SOP-22.0, AUXILIARY FEEDWATER SYSTEM. ,...-------,--------------------------,-----------, FNP-I-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response n 14 [CAl Check if SI accumulators should be isolated. Response NOT Obtained NOTE: Step 14.1 is a continuing action. 14.1 [CA] Check at least two RCS hot leg temperatures -LESS THAN 350°F. RCS HOT LEG TEMP [] TR 413 14.2 Reset S1. [] MLB-l 1-1 not lit (A TRN) [] MLB-l 11-1 not lit (B TRN) 14.3 Close all SI accumulator discharge valves. lA(lB,lC) ACCUM DISCH ISO [] QIE21MOV8808A [] QIE21MOV8808B [] QIE21MOV8808C _Page Completed 14.1 Perform the following. 14.1.1 WHEN at least two RCS hot leg temperatures are less than 350°F, THEN perform steps 14.2 and 14.3 to isolate accumulators. 14.1.2 Proceed to step 15. 14.2 CAUTION: Do not mention S821 since part of SRO WE11EG2.1.28 OBSERVE CAUTION PRIOR TO STEP 15. II any train will NOT reset using the MCB SI RESET pushbuttons, THEN place the affected train S821 RESET switch to RESET. ( SSP S TE S T CAB.) 14.3 Perform the following. 14.3.1 Vent any SI accumulator that cannot be isolated. ACCUM N2 VENT [] HIK 936 open lA(lB,lC) ACCUM N2 SUPP/VT ISO lA QIE2lHV [] 887 SA [] 887sB [] open open 14.3.2 IF an accumulator can NOT be isolated or vented, THEN consult the TSC staff to determine contingency actions. Page 18 of 22 , (]) E i-(j) ( \ -co 'r" -(j) (]) -.. co o co U N I.() o I'-'r" o -

100 I.

t.5 $;oOW!<.!

u-. "\J'.I .. c; .. OHO P'Ce\;r"',o,,", , !=OI't MO'tOR. "NO SIQ ONG>

s,y Date: 10/23/2009 4$) I_ 0PaN '\_ J(G, .. ... UI -1f-.a:o*o:O) <>il-<>SP 0; \-<> flP +SPo;jf-oSP 011-<> Sja "'-II-oil' <>If<> ISP o;jf-o!to .!!:!. 17 ... Title: G:\D177052.cal Time 09:36:33 AM -I I Enlarged view

53. 075A2.03 001/N8N/RO/C/A 2.5/2.7/N/N/4/CVR/Y Unit 1 is at 20% power, and the following conditions occurred:

At 1000:

Main Condenser pressure is 1.3 pSia and degrading.
AOP-8.0, Partial Loss Of Condenser Vacuum, is in progress due to an air ejector malfunction.

At 1010:

Main Condenser vacuum has degraded to 12 psia.
AOP-13.0, Condensate And Feedwater Malfunction, has been entered. Which one of the following describes the Circulating Water (CW) outlet temperature at 1010 as compared to earlier, and the action(s) required by AOP-13.0?

At 1010 CW outlet temperature is (1) than at 1000, and AOP-13.0 requires (2) (1 ) (2) A. higher tripping the reactor B. lower reducing power to approximately 2% C. higher reducing power to approximately 2% D':" lower tripping the reactor ------A -Incorrect. The first part is incorrect, since the Steam dumps are not able to arm with vacuum worse than 8" Hg Vacuum (10.78 psia), the SGFPs trip at 5.9 psia (12 in Hga), and the Main Turbine trips at 21" Hg (4.41 psia), Plausible, since it would be correct if Steam dumps armed and operated as usual. The second part is correct (see D). Plausible, even when combined with the first part, since the Main Condenser pressure at which the Control interlock C-9 for blocking Steam Dump operation on high pressure (lOW vacuum) is higher (a lower vacuum) than for a trip of the SGFPs and Main Turbine. B -Incorrect. The first part is correct (see D). The second part is incorrect, since AOP-13 directs tripping the reactor with a loss of both SGFPs >5%. Plausible, since the choice would be correct if initial power was 5% or less per AOP-13.0 Step 2.2 AlER "Reduce reactor power to approximately 2%". AOP-13.0 until recently allowed reducing power to the capacity of AFW for both SGFPs tripped from a power level as high as 35% to prevent a reactor trip. C -Incorrect. The first part is incorrect (see A). The second part is incorrect (see B). D -Correct. The first part is correct, since no steam is condensing in the Main Page: 142 of 2n 12/14/2009 Condenser from the Main Turbine, SGFPs, or from the Steam dumps. Due to the degraded vacuum, the steam dumps don't arm or operate, the Main Turbine Trips, the SGFPs trip, and the CW outlet temperature approaches the inlet (decreases from it's value when steam was condensing in the condenser). The SG atmospherics and Safeties will open as necessary to maintain SG pressure less than the design, and thus control Tavg. The second part is correct per AOP-13.0 step 2.1: for power above 5%, with a trip of both SGFPs, trip the reactor. Per FNP-O-SOP-0.3, Version 39.0, APPENDIX G: 8" Hg V cc. Mini mum to arm SDs, and 10.78 psi a max pressure to arm SDs. Per ARP-1.10, KK1, TURB COND VAC LO alarm, Version 64.0: NOTE: IF condenser vcnJum docra3S8Sto 21" Hg (4.41 psi a) , THEN aturbinetrip occurs. Per AOP-8.0, Partial LossOf Condenser Vacuum, Version 24.0: NOTE:

Main turbinetrip will occur at 4.41 psi a (9 in Hga)
SGFPtrip will occur at 5.9 psi a (12 in Hga). Procedures:

AOP-3.0 AOP-3.0, & AOP-13.0 AOP-3.0, AOP-13.0, & EEP-O FNP-1-AOP-13.0, Condensate And Feedwater Malfunction, Version 29.0 2 A/ER: Check Both SGFPs-TRI PPED 2 RNO: Proceed to step 3 OBSERVE CAUTION prior to step 3. 2.1 A/ER Check Recdor Power -LESS THAN 5%. 2.1 RNO: Trip the recdor and go to FNP-1-EEP-0, REACTOR TRIPOR SA.FETY INJECTION. 2.2 A/ER Reduce recdor power to approxi mately 2%. FNP-1-AOP-3.0, Turbine Trip Below P-9 Set:point, Version 16.0 FNP-O-SOP-0.3, Version 39.0, APPENDIX G, 8" Hg V cc. Mini mum to arm SDs, 10.78 psi a max pressure to arm SDs AOP-8.0, Partial LossOf Condenser Vacuum, Version 24.0

Main turbinetrip will occur at 4.41 psi a (9 in Hga)
SGFPtrip will occur at 5.9 psi a (12 in Hga) Page: 143 c:I 277 12/14/2009 Previous NRC exam history if any: 075A2.03 075 Ci rcul ati ng Water System A2 Ability to (a) predict the impactscl thefoilONing malfunctioosor operatioosoo thedrculating water system; and (b) based 00 thoee predictions, use procedurestocorrect, control, or mitigate the consequences cI those malfunctions or operatioos (CFR: 41.5/43.5/45.3/45.13)

A2.03 Safety featurescnd reiatiom:t1ip betw*EJ1 COnd6193r vocuum, turbinetrip, and steam dump .. 2.52.7* Match justification: This question presents a loss of vacuum and requires knowledge of how this affects the Circ Water system in this set of conditions: i.e. at this vacuum, the steam dumps are prevented from opening to protect the main condenser from overpressure. This causes CW temp to be affected differently than if they opened after the turbine trip as they normally would. The second part of the question and each choice requires knowledge of the procedure action for this condition (on the RO level). Objective:

3. STATE AND EXPLAIN the operational implications for all Cautions, Notes, and Actions C5OOCiatoo with AOP-B.O, Partial Loss of Condalser Vccuum. (OPS-52520H03).

5. EVALUATE plant conditions and DETERM I NE ifc:ny 5)'stem componaltsnero to be operatoo whi I e performi ng A OP-B.O, Parti aI Loss of Condalser V ccuum. (OPS-52520H06) . Page: 144 cl2n 12/14/2009 Question # 53 KIA 075A2.03 REFERENCE Docs 04/03/09 13:20:52 FNP-I-AOP-8.0 PARTIAL LOSS OF CONDENSER VACUUM Version 20.0 I I! Action/Expected Response Response Not Obtained NOTE:

DEHC point CNDPI displays condenser pressure in Hga absolute on the point detail page. On all other pages CNDP I displays condenser pressure in psia. 1 2
IPC points PT0214 and PT0215 display condenser pressure in psia.
MCB recorder PR 4029 displays condenser pressure in psia.
Main turbine trip will occur at 4.41 psia (9 in Hga) < p-9 therefore No Rx Trip on turbine trip.
SGFP trip will occur at 5.9 psia (12 in Hga) Monitor Condenser pressure 1 .1 IF condenser pressure is at 2.3 psia, THEN increased monitoring of condenser pressure is required.

Monitor turbine trip criteria. 2.1 WHEN turbine power greater than or equal to 30%, THEN verify condenser pressure less than 2.7 psia (5.5 in Hga). LOSS of All MFW pumps--> requires a trip per AOP-13 when> 5%. 2.1 Perform the following. 2.1.1 IF reactor power greater than 35%, THEN trip the reactor. 2.1.1.1 Go to FNP-I-EEP-O, REACTOR TRIP OR SAFETY INJECTION. 2.1.1.2 Perform the remainder of this procedure in conjunction with FNP-I-EEP-O, REACTOR TRIP OR SAFETY INJECTION. 2.1.2 IF reactor power less than 35%, THEN place MAIN TURB EMERG TRIP to TRIP for:::: 5 seconds. 2.1.2.1 Perform FNP-I-AOP-3.0, TURBINE TRIP BELOW P-9 SETPOINT in parallel with this procedure. 2.1.3 Proceed to step 3. o Step 2 continued on next page Page Completed Page 2 of9 04/03/0913:21:19 FNP-I-AOP-13.0 CONDENSA TE AND FEEDWA TER MALFUNCTION Version 29.0 I I I Action/Expected Response 2 1.13 Check reactor power change < 15% 1.14 Check parameters within limits for continued at power operation.

Pressurizer level greater than 15%
Pressurizer pressure greater than 2100 psig
SG narrow range levels 35%-75%
Control rod bank position Lo-Lo Annunciator FE2 Clear
Delta I within limits specified in the COLR Check Both SGFPs -TRIPPED 2.1 Check Reactor Power LESS THAN 5%. 2.2 Reduce reactor power to approximately 2 Response Not Obtained 1.12.5 Place STM DUMP INTLK TRAIN A and B to ON. 1.13 Notify Shift Radiochemist to sample the RCS per FNP-I-STP-746.

1.14 IF the Team is NOT confident that a parameter is being restored, THEN trip the reactor and go to FNP-I-EEP-O, REACTOR TRIP OR SAFETY INJECTION. Correct part #1 Proceed to step 3 OBSERVE CAUTION prior to step 3. 2.1 Trip the reactor and go to FNP-I-EEP-O, REACTOR TRIP OR SAFETY INJECTION. 2%. B & C distractors part #2 2.2.1 2.2.2 2.2.3 Verify rod control in MANUAL. Stabilize T A VG by adjusting rod position and/or boron concentration. Check for proper operation of steam dumps. o Step 2 continued on next page Page Completed Page 6 of23 04103109 13: 17 :30 Interlock Source Setpoint 6. C-7 Turb. Impulse 15% Turb. Sudden Loss Press Instr. 447 Power Reduction of Load rate ckt. 120 sec Time Const. 7. C-9 Condo Press. 8" Hg Vac. and Condenser Switch and Circ Closed Available Water Pump Bkrs. 8. C-ll PI A Converter 220 Steps BankD Stop BankD Position 9. C-20 Turb. Impulse >40% AM SAC Pressure Inst. Enabled 2446 and 2447 FNP-0-SOP-0.3 APPENDIX G CONTROL INTERLOCKS Coincidence & Light Status Function 111 > setpoint Arms steam dump valves. lit> setpoint Manually reset by placing the control mode selector switch to reset momentarily. 2/2 > setpoint Allows steam dump valves to be armed. and 112 > setpoint lit > setpoint NO stm dumps 111 = setpoint Stops outward rod motion in auto. lit after being at setpoint for 3 min. 2/2 > setpoint; Allows AMSAC to be armed. also sealed in for 260 seconds after Pimp lowers to < 40%; lit < setpoint Page 70f7 Version 38.0 07/02/0906:30:47 SETPOINT: I. 1.3 psia when < 30% power 2. 2.3 psia when> 30% power ORIGIN: OEH PROBABLE CAUSE FNP-I-ARP-I.IO LOCATION KK I TURB CONOVAC LO I. Low Circulating Water Flow or I Turbine trip and no dumps results in significant

2. Gland Seal Steam System fault. reduction in Mass flow into the condenser

3. Improper valve lineup resulting i AUTOMATIC NOTE: IF condenser vacuum decreases to 21" Hg (4.41 psia), THEN a turbine trip occurs. q umd m slm q cOl1 q mild ,J.. ,.,!..= m cir(1fa(er m cimralel' C B NONE c (r out -T 111 ),J.. (;rill)B OPERATOR ACTION (Toul -Tin ),J..=> (Tom),J..

,-,au \IV me ml"ume is operating at 30% load, THEN the maximum permissible condenser pressure is 5.5 inches Hg (2.7 psia). WHEN the turbine is operating at < 30% load, THEN the maximum permissible condenser pressure is 3.5 inches Hg (1.7 psia). I. Perform the actions required by FNP-I-AOP-8.0, PARTIAL LOSS OF CONDENSER VACUUM.
References:
A-I77100, Sh. 491; 0-172803; 0-170812, Sh. 2; U-162213, Tab 5; Westinghouse Customer Advisory Letter 86-02; OCP P-95-1-8943 Page 1 of 1 Version 62.0
54. 076A3.02 001/NEW/RO/M EM 3.7/3.7//N/3/CVR/Y A Safety Injection and loss of B train Start Up transformer occurred on Unit 2. Which one of the following is the position of the Turbine Building Service Water Supply Isolation Valves? Valve nomenclature is listed below: SW TO TURB BLDG ISO B TRN V514 SW TO TURB BLDG ISO A TRN V515 SW TO TURB BLDG ISO A TRN V516 SW TO TURB BLDG ISO B TRN V517 V515N517 V514N516 A. Throttled Throttled B. Closed Throttled C. Throttled Closed D'f Closed Closed Page: 145 of 277 12/14/2009 A -Incorrect.
This would be correct for LOSP on both trains with no SI, but in this case there is an SI on both trains and an LOSP on only one train. B -Incorrect. The first part is correct because of the SI on A train. The second part is incorrect since the SI isolates both trains of valves regardless of the LOSP signal on the B train. This would be correct for an LOSP on B Train with no SI, and an SI on A train with or without an LOSP on A train, but in this case the SI isolates both trains of SW valves. C -Incorrect. This is the exact opposite of B and may be chosen if confusion existed as to the automatic action of the SW to the TB valves due to the two signals: SI and LOSP. D -Correct. The SI Closes the valve on both trains, even though the LOSP alone (with no SI) would throttle the valves on the respective train. With an SI and and LOSP, the valves close to ensure sufficient cooling flow to the Emergency DGs. SW FSD A-181 001 3.44 TURBINE BUILDING SERVICE WATER SUPPLY ISOLATION VALVES TPNS Nos. Unit 1 Unit 2 Train A -Q1 P16V515 Q2P16V515 Q1 P16V516 Q2P16V516 Train B -Q1P16V514 Q2P16V514 Q1 P16V517 Q2P16V517 3.44.1 Basic Functions Redundant Turbine Building Service Water Supply Isolation Valves automatically isolate the nonessential turbine building service water loads upon receipt of a Phase A Containment Isolation Signal (T-signal) and/or excess turbine building service water flow rate. This action is required to ensure adequate service water flow to safety-related equipment during accident modes. The Turbine Building Service Water Supply Valves provide a second, throttling function during a Loss of Offsite Power event. Specifically, the valve operators automatically position the valve to 16 degrees in the open direction upon receipt of a LOSP signal. This throttled, or mid-stroke, position serves to provide a limited amount of cooling water to the Turbine Bldg. to support the cooldown of the secondary side of the plant. This action simultaneously serves to automatically provide increased cooling water to the Emergency Diesel Generators during the LOSP event. Plant operator actions are still required to isolate the Turbine Building within fifteen minutes to provide cooling water for long term operation of the diesels. Page: 146 of 277 12/14/2009 Previous NRC exam history if any: 076A3.02 076 Sevi ce WatEr SystEm A3 Ability to monita autanaticcperation of theSWS, including: (CFR: 41.7/45.5) A3.02 Emergency heat 10008 ............................................ 3.73.7 Match justification: Ability to monitor automatic operations of the Service Water system including: emergency heat loads. This question requires knowledge of automatic operation of the SW system in an emergency as it automatically operates to reduce SW to non-vital loads to conserve SW for cooling the emergency DGs by throttling on an LOSP, and isolating the TB SW Supply on an SI. Objective:
6. ANALYZE plant conditions and DETERM I NE the successful completion of any step in AOP-10.0, Loss of SeviceWatEr. (OPS-52520J07).
Page: 147 of 277 12/14/2009 Question # 54 KIA 076A3.02 REFERENCE Docs FNP Units 1 & 2 3.43.4 3.43.5 SERVICE WATER SYSTEM A-18lO01 These pressure switches shall meet or exceed seismic qualification requirements contained in IEEE 344-1975. (References 6.7.023 and 6.7.024) Interface Requirements The Electrical Distribution System shall provide power through the Annunciator Cabinet I C for audible and visual annunciation upon closing of the pressure switch contacts. (References 6.4.202 -6.4.205) Failure Modes and Effects Analysis There are redundant pressure switches and low pressure alarms on each train of service water. Therefore, should SW pressure approach the low pressure setpoint, the failure of a single pressure switch will not prevent an alarm. A spurious low pressure alarm from one pressure switch could be checked for validity against the Auxiliary Building header pressure indicator. If the alarm appears true, plant operators should check the SW pumps for proper operation or take other actions as appropriate. 3.44 TURBINE BUILDING SERVICE WATER SUPPLY ISOLATION VALVES TPNS Nos. Train A-Train B -3.44.1 a,b,c plausibility Unit 1 Unit 2 QIP16V515 Q2PI6V515 QIP16V516 Q2Pl6V516 QIP16V514 Q2P 16V514 QIP16V517 Q2P16V517 Basic Functions Redundant Turbine Building Service Water Supply Isolation Valves automatically isolate the nonessential turbine building service water loads upon receipt of a Phase A Containment Isolation Signal (T -signal) and/or excess turbine building service water flow rate. This action is required to ensure adequate service water flow to safety-related equipment during accident modes. The Turbine Building Service Water Supply Valves provide a second, throttling function during a Loss of Offsite Power event. Specifically, the valve operators automatically position the valve to 16 degrees in the open direction upon receipt of a LOSP signal. This throttled, or mid-stroke, 3-103 Ver.46.0 FNP Units 1 & 2 3.44.2 3.44.3 SERVICE WATER SYSTEM A-18l00l position serves to provide a limited amount of cooling water to the Turbine Bldg. to support the cooldown of the secondary side of the plant. This action simultaneously serves to automatically provide increased cooling water to the Emergency Diesel Generators during the LOSP event. Plant operator actions are still required to isolate the Turbine Building within fifteen minutes to provide cooling water for longterm operation of the diesels. Functional Requirements 3.44.2.1 3.44.2.2 3.44.2.3 3.44.2.4 3.44.2.5 3.44.2.6 The valve design inlet pressure and temperature conditions must be 150 psig and 115°F. (References 6.5.007 and 6.7.153) The valve operator shall be capable of opening and closing the valve at a maximum differential pressure of 150 psig. (Reference 6.5.007) The Turbine Building Service Water Supply Isolation Valves shall be electric motor operated. A handwheel must be provided on each valve operator to stroke the valve in the event of motor failure. (Reference 6.5.007) The Unit 1 SW Supply Isolation Valves are not readily accessible for manual operation due to the welded security covers on Valve Box VB-I. The Inservice Testing Plan has established a stroke time for closing these motor operated valves of less than or equal to 75 seconds (+7.5 seconds). However, there are no system functional requirements imposing a specific stroke time. Motor operators shall be capable of starting and accelerating the load with 80% nameplate voltage at the motor terminals. (References 6.5.007, 6.7.123, 6.7.150, and 6.7.175) Safety-related MOVs are subject to the requirements of NRC Generic Letter 89-10, dated 6/28/89, "Safety-Related Motor Operated Valve Testing and Surveillance." (Reference 6.7.043) Code Requirements 3.44.3.1 The valves must be designed in accordance with requirements of ASME B&PV Code, Section III, Class 3, 1971 Edition along with additional design codes and 3-104 Ver. 46.0
55. 077AA 1.02 001/BANK/RO/M EM 3.8/3.7/N/N/3/HBF/Y The following plant conditions exist on Unit 1:
100% power.

All systems are aligned normally.

Generator reactive load is currently at "0" MVARs.

ACC has notified the plant that system voltage problems require UNIT 1 to establish maximum allowable incoming reactive load (MVARS in). Which one of the following:
1) identifies the administrative limit on incoming reactive load (MVARS in) lAW UOP-3.1, Power Operation, and 2) the proper switch which will be used to establish maximum allowable incoming reactive load? (1 ) (2) A. -200 MVARs Manual Voltage Adjust Switch B. -200 MVARs Auto Voltage Adjust Switch C. -300 MVARs Manual Voltage Adjust Switch D':" -300 MVARs Auto Voltage Adjust Switch Page: 148 c:J 277 12/14/2009 A -Incorrect.

1) -200 MVARS is the limit of the MCB MVAR meter; UOP-3.1 step 3.3.3 states that -300 MVAR is the administrative limit. Also stated in SOP-36.8 4.8.2.2. Plausible:
This is a limit of the MCB meter, and could be perceived to be the administrative limit for operation.
2) Operation of the Manual voltage adjust is incorrect per SOP-28.1 , see caution before 4.7.14 and 4.23.1 ; operating the Manual voltage adjust while in auto changes the base for Auto and if it were to auto shift to manual, the transient could result in generator damage/rx trip. Plausible:
A manual adjustment is being made, and the Manual Voltage Adjustment Switch is manipulated if the voltage regulator were in TEST or OFF; this switch is also manipulated to ZERO VM4098. B -Incorrect.
1) See A for discussion and plausibility.

2) this is the correct switch to manipulate. C -Incorrect.

1) this is the correct Administrative limit per UOP-3.1 (ver 101) 3.3.3 2) See A #2 for discussion and plausibility. D -Correct. 1) UOP-3.1, SOP-36.8 both establish this limit for MVARs to prevent the auto adjuster from going to its mechanical stop. 2) This is the correct switch manipulation per SOP-28.1 for adjusting MVARS. UOP-3.1, Version 104.0 SOP-36.S, Version 14.0 Page: 149 of 2n 12/14/2009 Previous NRC exam history if any: Sequoyah 2009 question 17 (RO NRC EXAM) 077AA1.02 077 Generator Voltage and Electric Grid Disturbances AA1. 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 ) AA1.02 Turbine / generator controls .......................................................
3.83.7 Match justification:
Grid disturbance:
identified in stem, causing the initiating queues for the operator to "OPERATE and/or MONITOR Turbine/generator controls--- Grid disturbances result in Alabama Control Center (ACC) or Power Coordination Center (PCC) to request FNP to accept max vars in.
The examinee must know limitations to the operation of the voltage regulator and select the correct switch to manipulate the Generator output. Objective:
OPS-52105C05; Recall and discuss the P&Ls, Notes, and Cautions found in SOP-28.1. OPS-40202B16; State and explain how the [ ... ] main generator parameters are controlled, including thier limits, and the adverse effects on the main generator of excessive VARS OUT or VARS IN. Page: 150 of 277 12/14/2009 Question # 55 KJA 077AA1.02 REFERENCE Docs 07/02/0906:39:10 FNP-I-UOP-3.1 3.3 Generator and Switchyard 3.3.1 Do not adjust generator excitation with the Manual Voltage adjust switch when the generator is on the line without Operations Manager approval. 3.3.2 WHEN adjusting VARs, THEN do NOT exceed 22 kv +/- 5% (20.9 -23.1 kv). IF outside this band, THEN reduce generator output due to stator overheating. 3.3.3 The Admin. limit for main generator MVARs is -300 to prevent the auto adjuster from going to its mechanical stop. 3.3.4 IF plant conditions are met per APPENDIX 5, LOAD LIMITATIONS WITH A 500 Kv TRANSMISSION LINE OUT OF SERVICE, THEN the total plant MW limitation value has to be met within 30 minutes from the initiating condition(s). (AI 2004204794) 3.3.5 Alabama Control Center should be notified any time a Power System Stabilizer is removed from or returned to service. 3.3.6 The NERC standards in APPENDIX 6 will be met whenever the plant is on line. These standards were promulgated in accordance with the Energy Policy Act to ensure grid reliability. 3.4 Chemistry 3.4.1 Maintain the RCS and steam generator chemistry within limits required by FNP-0-CCP-202, CHEMICAL SPECIFICATIONS. 3.4.2 Chemistry should be notified of the following:
To sample RCS per STP-746 ifRx power changes by:::: 15% of rated thermal power within a 1 hr period. (SR 3.4.16.2)

Any significant changes in plant load Version 101.0 09118/09 15 :39:54 FNP-0-SOP-36.8 Quality Guide, and FSAR voltages for the ESF components.
Ensure an Auto Log entry is made similar to the following: "During the Unit 112 refueling/forced outage, voltage will not be able to be raised / lowered to meet the APC 500kV voltage schedule. Relief granted by ACC (System Operator name)." 4.8.1.6 Unit capability curves should be adhered to when attempting to maintain bus voltage schedule. 4.8.1. 7 Station service voltage limits should be observed when attempting to maintain bus voltage. 4.8.1.9 The Voltage is subject to change at the request of the ACC System Operator. ACC System Operator requests shall be met in a timely manner. Any emergency request shall be met as soon as possible. 4.8.1.10 The System Operator would like to minimize switching the shunt reactor in/out of service as much as possible. With one Unit off for refueling, placing the shunt in service, and leaving it in service, appears to be successful in allowing Farley to maintain the Voltage Schedule within the NERC allowable, undirected deviation. This also keeps the shunt reactor switching to a minimum. 4.8.1.11 The ACC System Operator does not want to deviate too far from the scheduled voltage because a single contingency could put Farley in a situation for high / low on-site voltages, and possible damage to equipment. 4.8.1.12 The ACC System Operator is allowed some discretion for short durations from the allowable, undirected deviation from the Voltage Schedule. 4.8.2 Guidelines to Raise and Lower System Voltage The Farley Operators shall adjust the main generator reactive load voltage to meet the system requirements as directed by the System Operator, while observing the following guidelines. 4.8.2.1 4.8.2.2 4.8.2.3 The reactive load adjustments cannot exceed 22kV +/-5% (20.9 -23.lkV). The Farley administrative limit is -300 MV ARs to prevent the auto adjuster from going to its mechanical stop. The 230kV Shunt Reactor is placed in service when the 230kV bus voltage needs to be lowered, and the 230kV Capacitor Bank is placed in service when the 230kV bus voltage needs to be raised. Because the two devices perform opposite functions, they never should be in service at the same time, and an interlock scheme is provided on switches 955 and 957 to prevent this from happening. Page 25 of 27 Version 14.0 07/02/0906:34:49 FNP-J-SOP-28.1 4.23 Generator Voltage Regulator Balance Meter Null Adjustment When Operating In Auto Voltage Control CAUTION: DO NOT use manual voltage adjust switch to adjust generator voltage when in auto voltage control. Use the auto voltage adjust switch to adjust generator voltage, and the manual voltage adjust switch to adjust voltage regulator balance (null) meter. NOTE: The voltage regulator balance meter should be monitored more frequently during load changes. 4.23.1 4.23.2 4.23.3 Monitor the voltage regulator balance meter to ensure it remains as close to zero as practical (approximately +/-0.25 volts). IF null meter reading less than zero, THEN raise voltage regulator balance meter indication by slowly taking Manual Voltage Adjust Switch to lower. IF null meter reading greater than zero, THEN lower voltage regulator balance meter indication by slowly taking Manual Voltage Adjust Switch to raise. Version 110.0 07/02/0906:34:49 FNP-I-SOP-28.1 NOTE: 4.7.9 4.7.10 Have Maintenance check that no trip signal is present on the line distance relay, NIN3IRLYGEN2IKD. (AI 2008207695) 4.7.9.l 4.7.9.2 IF no trip signal is present, THEN reconnect the output of the relay by closing knife switch labeled "GENERATOR OVERCURRENT TRIP (KD-41 LOCKOUT)" located on Meter & Relay Panel No.6 (N 1 H 11 L506). IF a trip signal is present, THEN write a condition report to investigate the reason for a trip signal before reconnecting the output from the relay. Using the Manual Voltage Adjuster Switch, increase generator output voltage to 22 KV, +/- 0.2 KV, as indicated on analog voltmeter 4099 or digital voltmeter 5122. (CR20091 03418) In the following step, the amber bar will be lit while the Regulator Transfer Switch is in the TEST osition. CR2009103418 4.7.11 Place the Regulator Transfer Switch in the TEST position. NOTE: Raising the voltage with the Auto Voltage Adjust Switch will cause the null meter (VM4098) to move in the positive direction. CAUTION: CAUTION: 4.7.12 Using the Automatic Voltage Adjuster Switch, null (zero) the regulator balance voltmeter. To prevent a possible turbine trip that could occur if the regulator does not null out properly, perform step 4.7.14 without delay after completing step 4.7.13. 4.7.13 Place the Regulator Transfer Switch in the ON position. All. generator voltage adjustments must now be made with the Automatic Voltage Adjuster Switch. 4.7.14 Verify generator output voltage is approximately 22KV as indicated on analog voltmeter 4099 or digital voltmeter 5122. Adjust voltage as necessary using the Automatic Voltage Adjuster Switch. Version 110.0 07/02/09 06:34:49 APPENDIXB FNP-I-S0P-28.1 APPENDIX B TRANSFERRING THE VOLTAGE REGULATOR TO MANUAL 1.0 Have I&C attach a digital voltmeter across Voltage Regulator Transfer Meter VM4098. NOTE: IF the Regulator Transfer Meter VM4098 needle has slight oscillations, THEN the meter should be nulled based on the average or midpoint of the oscillations. 2.0 Zero the Voltage Regulator Transfer Meter VM4098 using the Man Voltage Adjust Switch using the digital voltmeter for closer match. 2.1 For 0 to 150 MVARs IN or for 0 to 100 MVARs OUT, a slightly positive value (between 0 & +0.1 V) should be obtained on the Voltage Regulator Transfer Meter VM4098 using the Manual Voltage Adj Switch. 2.2 For 100 to 300 MVARs OUT, a slightly negative value (between 0 & -0.1 V) should be obtained on the Voltage Regulator Transfer Meter VM4098 using the Manual Voltage Adjust Switch. 2.3 For more than 150 MVARs IN or more than 300 MVARs OUT, transfer to Manual should not take place without further evaluation unless an emergency condition exists. In the event of an emergency, use the guidance in step 2.1 to null for> 150 MVARs IN, and use the guidance in step 2.2 to null for> 300 MVARs OUT. 3.0 Going to Manual on the Voltage Regulator NOTE: WHEN the voltage regulator is turned to TEST or OFF, THEN the PSS inputs are disabled. 3.1 Turn Voltage Regulator Transfer Switch to TEST. 3.2 IF a large swing in generator voltage occurs, THEN turn Voltage Regulator Transfer Switch back to ON. 3.3 IF staying in manual on the Voltage Regulator, THEN record in the RO's Log the date, time, and reason for going to manual on the Voltage Regulator. (NERC Phase III) NOTE: WHEN the voltage regulator is turned to TEST or OFF, THEN the PSS inputs are disabled. 4.0 If required the Voltage Regulator Transfer Switch may be turned to OFF. Page I of2 Version 110.0
1/2009 Sequoyah RO NRC Exam 21412009 17. Given the following: Unit 1 is at 100% RTP. -All systems are aligned normally. -Generator reactive load is currently at "0" MV ARs. -The Transmission Operator has notified the plant that system voltage problems require Unit 1 to establish the maximum allowable outgoing reactive load. Which ONE of the following identifies the MAXIMUM outgoing reactive load in accordance with GOI-6, "Apparatus Operation", and the correct operation of the Exciter Voltage Adjuster? Maximum Outgoing Reactive Load Exciter Voltage Adjuster A. 240 MVARs Lower B. 240 MVARs Raise C. 300 MVARs Lower D. 300 MVARs Raise Page 17
56. 078K2.02 001/NEW/RO/M EM 3.3/3.5/N/N/2/CVR/Y Which one of the following correctly states the power supplies to the 1 A and 1 B Emergency Air Compressors?
A'! 1 A and 1 B 600V MCC B. 1A and 1 B 600V LCC C. 1 D and 1 E 600V LCC D. 1 U and 1V 600V MCC A -Correct. Per the load list A-506250, Rev. 12, Pages F -94 & G -78. B -Incorrect. Plausible, since the voltage is correct and these are safety related switchgears. C -Incorrect. Plausible, since the voltage is correct and these are safety related switchgears. D -Incorrect. Plausible, since the voltage and type of switchgear is correct: safety related 600V Motor Control Center (MCC). LOAD LIST, A-506250, Rev. 12 -94& G -78 Previous NRC exam history if any: 078K2.02 078 I nstrument Air System K2 KnONlecigeof buspoNer SJppli es to the following: (CFR: 41.7) K2.02 ElTla'gency ai r compresoor ........................................ 3.3* 3.5* Match justification: FNP has 2 Emergency Air Compressors which Air to operate SG Atmospheric Relief valves if Instrument air is unavailable. This question requires knowledge of the bus power supplies for the two Emergency air compressors. Objective: 1 NAM E AND I DENTI FY the Bus power supplies, for those electrical components associatEd with the CompressEd Air System, to include those items in Table 1-Power Suppl i as (OPS-40204D04). Page: 151 of 2n 12/14/2009 Question # 56 KIA 078K2.02 REFERENCE Docs FNP UNIT I DG03 EE10 1B 600/208V MCC (CONT'D) BKR TPNS --FBI6 ------------ FBJ2L ------------ FBJ3L ------------ FBJ4R ------------ FBJ5L Q1R22LOOO6B-B FBJ5R Q1P1SMOOO2B-B FBJ7 N1T49MOOO1B-N FBK2 ------------ FBK5L QIR17BOOO2-B FBM2 ------------ FBM3 Q1P12MOOO1B-B FBM4L N1E15KOOO2A-N FBM5 Q1E16MOOO1B-AB FB02 N1T49MOV3310B-N FB03 ------------ FB04 ------------ Isectg.doc LOAD LIST A-506250 AB -121' B177556-2 DESCRIPTION SEE PAGE SPARE SPARE SPARE SPARE I 3-15 KVA SINGLE PHASE CONSTANT VOLTAGE G-SO TRANSFORMERS >>> 1H REG AC DIST PNL >>> 1B EMERG AIR COMP FOR MAIN STEAM ATMOS RELIEF VALVES 1B REACTOR CAVITY COOLING FAN SPARE 1B 600/208V MCC XF.MR >>> IB MCC 208V G-S2 SECTION SPARE 1B REACTOR MAKEUP WATER PUMP PENETRATION FILTER MECH EQPT ROOM UNIT HEATER 1A DISC SWITCH Q1R1SB003B-B >>> 1B CHARGING/HHSI PUMP ROOM COOLER FAN REACTOR CAVITY COOL FAN MOV SPARE SPARE Page G -78 Ver.54.0 FNPUNIT I DF03 EDI0 lA 600/208V MCC (CONT'D) BKR TPNS FAJ5 Q1E19MOOO1C-A FAJ6 -------------- FAJ7 -------------- FAK2L Q1R37EOOO1A-A Q1R37LOOO4A-A FAK2R -------------- FAK4L N1DllREOO1O-N FAK4R -------------- FAK6 -------------- FAL2 -------------- FAL3L Q1E17GOOO1A-A FAL3R -------------- FAL5L Q1P18MOOO2A-A FAM4 -------------- FAM5L N1G21NDRE2608-N FAM5R -------------- lsectf.doc LOAD LIST A-S062S0 AB -139' B177556-1 DESCRIPTION SEE PAGE lC CTMT POST LOCA AIR MIXING FAN SPARE SPARE TRANSFORMER lA (Q1R37EOOO1A-A) >>> lB-A SIS DISTRIBUTION PANEL-TRAIN A Q1R37LOOO4A-A (EXCEPT FOR BREAKER 24, ALL BREAKERS IN PANEL lB-A LEFT IN OPEN POSITION) >>> POST ACCIDENT HYDROGEN ANALYZER HEAT TRACING ALARM SPARE PENETRATION RM AIR PARTICLE DET MONITOR SPARE SPARE SPARE lA CTMT H2 RECOMBINER HEATER SPARE lA EMERG. AIR COMPRESSOR FOR MAIN STEAM ATMOS. RELIEF VALVES SPARE RECYCLE EVAP CONTROL PANEL I 600/208V MCC XFMR >>> lA MCC 208V SECTION F-97 Page F -94 Rev. 12
57. 103A4.03 001/MOD/RO/MEM 2.7/2.7/N/N/3/HBFNER 5 EDITORIAL The following plant conditions exist on Unit 1 :
A LOCA has occurred.

Containment Pressure is 30 psig and decreasing.

All required actuations have occurred.
Which one of the following describes the MINIMUM conditions if any, AND actions required to reset 'B' train PHASE B CTMT ISO (MLB-3 6-1)? A. 1) Containment Pressure must be lowered to less than the HI-3 setpoint prior to reset. 2) BOTH Train A and B CS RESET push buttons, and BOTH Train A and B PHASE B CTMT ISO RESET push buttons must be depressed. B. 1) Containment Pressure must be lowered to less than the HI-3 setpoint prior to reset. 2) The B Train PHASE B CTMT ISO RESET pushbutton ONLY must be depressed. 1) Phase B can be reset regardless of Containment Pressure.
2) The B Train PHASE B CTMT ISO RESET pushbutton ONLY must be depressed.
D. 1) Phase B can be reset regardless of Containment Pressure.
2) BOTH Train A and B CS RESET push buttons, and BOTH Train A and B PHASE B CTMT ISO RESET pushbuttons must be depressed.
Page: 152 of 277 12/14/2009 A Incorrect.
1) Phase B is equipped with a memory retentive latching relay when actuated on HIGH-3 signal. This latching relay allows for a reset of the signal before clearing the initiating signal. 2) Containment Isolation Phase B signal does not require Cnmt Spray signal to be RESET. These signals, because of the memory retentative relays are mutually exclusive of one another for RESET. plausibility:

1) SI, Phase A, and Phase B components can not be repositioned without first clearing the originating signal; as is true for the TDAFW pump UV & LO-Level auto-start signals. 2) CNMT Spray actuation requires the operation of two switches per train; Also, the Phase Band Cnmt Spray actuation signals are actuated by the same signal and it is feasible that one might believe that one must be reset before the other. B Incorrect.

1) See A part 1 for discussion and plausibility

2) See C part 2 for discussion.
C Correct. 1) The latching relays allow resetting the Phase B signal without clearing the actuating condition. See A part 1 for discussion.
2) This is correct. Depressing a single RESET pushbutton for EACH train will reset the PHASE B containment isolation signals, AND the B train pushbutton resets the B train signal. D Incorrect.

1) See C part 1 2) See A part 2. Previous NRC exam history if any: 103A4.03 103 Containment System A4 Ability to manually operate and/or monitor in the control room: (CFR: 41.7/ 45.5 to 45.8) A4.03 ESF slave relays ................................................ . 2.7* 2.7* Match justification:
Operation of ESF slave relays --resetting Phase B signal by depressing the RESET push buttons on the MCB, Depressing these pushbuttons resets the sealed in signal on the CNMT ISOL valves allowing them to be repositioned to support recovery actions. Objective: 52201107: Recall and describe the operation and function of the following [ ... ] ESF to include setpoint, [ ... ] and reset features [ .. .]. Page: 153 of 277 12/14/2009 Question # 57 KIA 103A4.03 REFERENCE Docs 3.3 INSTRUMENTATION Containment Purge and Exhaust Isolation Instrumentation 3.3.6 3.3.6 Containment Purge and Exhaust Isolation Instrumentation LCO 3.3.6 The Containment Purge and Exhaust Isolation instrumentation for each Function in Table 3.3.6-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.6-1. ACTIONS ----------------------------------------------------------NOTE----------------------------------------------------------- Separate Condition entry is allowed for each Function. CONDITION A. One Required radiation monitoring channel inoperable. B. ------------NOTE------------ Only applicable in MODE 1, 2, 3, or 4. -------------------------------- One or more Functions with one or more manual or automatic actuation trains inoperable. QE. Required Action and associated Completion Time of Condition A not met. Farley Units 1 and 2 A.1 B.1 REQUIRED ACTION COMPLETION TIME Restore the affected 4 hours channel to OPERABLE status. Enter applicable Immediately Conditions and Required Actions of LCO 3.6.3, "Containment Isolation Valves," for containment purge and exhaust isolation valves made inoperable by isolation instrumentation. 3.3.6-1 Amendment No. 146 (Unit 1) Amendment No. 137 (Unit 2) ACTIONS CONDITION C. ------------NOTE------------ C.1 Only applicable during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment. One or more manual channel(s) inoperable. Two radiation monitoring channels inoperable. Required Action and associated Completion Time for Condition A not met. Farley Units 1 and 2 C.2 Containment Purge and Exhaust Isolation Instrumentation 3.3.6 REQUIRED ACTION Place and maintain containment purge and exhaust valves in closed position. Enter applicable Conditions and Required Actions of LCO 3.9.3, "Containment Penetrations," for containment purge and exhaust isolation valves made inoperable by isolation instrumentation. COMPLETION TIME Immediately Immediately 3.3.6-2 Amendment No. 146 (Unit 1) Amendment No. 137 (Unit 2) Containment Purge and Exhaust Isolation Instrumentation 3.3.6 SURVEILLANCE REQUIREMENTS

NOT E --------------------------------------------------------- Refer to Table 3.3.6-1 to determine which SRs apply for each Containment Purge and Exhaust Isolation Function. SURVEILLANCE FREQUENCY SR 3.3.6.1 Perform CHANNEL CHECK. 12 hours SR 3.3.6.2 Perform ACTUATION LOGIC TEST. 92 days on a STAGGERED TEST BASIS SR 3.3.6.3 Perform MASTER RELAY TEST. 92 days on a STAGGERED TEST BASIS SR 3.3.6.4 Perform COT. 92 days SR 3.3.6.5 Perform SLAVE RELAY TEST. 18 months SR 3.3.6.6 -------------------------------- NOT E ------------------------------ Verification of setpoint is not required.

Perform T ADOT. 18 months SR 3.3.6.7 Perform CHANNEL CALIBRATION. 18 months SR 3.3.6.8 Verify ESF RESPONSE TIME within limit. 18 months on a STAGGERED TEST BASIS Farley Units 1 and 2 3.3.6-3 Amendment No. 180 (Unit 1) Amendment No. 173 (Unit 2) Containment Purge and Exhaust Isolation Instrumentation 3.3.6 Table 3.3.6-1 (page 1 of 1) Containment Purge and Exhaust Isolation Instrumentation FUNCTION 1. Manual Initiation
2. Automatic Actuation Logic and Actuation Relays 3. Containment Radiation Gaseous (R-24A, B) APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS 1,2,3,4, (a), (b) 1,2,3,4 1,2,3,4 (a), (b) REQUIRED CHANNELS 2 2 trains 1 2 SURVEILLANCE REQUIREMENTS SR 3.3.6.6 SR 3.3.6.2 SR 3.3.6.3 SR 3.3.6.5 SR 3.3.6.8 SR 3.3.6.1 SR 3.3.6.4 SR 3.3.6.7 TRIP SETPOINT NA NA -2 ,,; 2.27 X 10 IJCiicc (c)(d) -3 ,,; 4.54 X 10 IJCiicc (c)(e) -3 ,,; 2.27 X 10 IJCiicc (c)(f) 4. Containment Isolation
-Phase A Refer to LCO 3.3.2, "ESFAS Instrumentation," Function 3.a., for all initiation functions and requirements. (a) During CORE ALTERATIONS. (b) During movement of irradiated fuel assemblies within containment. (c) Above background with no flow. (d) With mini-purge in operation. (e) With slow speed main purge in operation. (f) With fast speed main purge in operation. Farley Units 1 and 2 3.3.6-4 Amendment No. 146 (Unit 1) Amendment No. 137 (Unit 2) NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Reactor Operator Question #27 Examination Outline reference: Level Tier # Group # KIA # Importance Rating Ability to manually operate and/or monitor in the control room: ESF slave relays Proposed Question: Common 27 Given the following conditions:
A LOCA has occurred.

Containment Pressure is 28 psig and rising.

All required actuations have occurred.
RO 2 1 103 A4.03 2.7 SRO Which ONE (1) of the following describes the conditions required and operation of relays to reset Containment Isolation Phase B? A. The Phase B slave relays are de-energized when the Phase B control switch is placed in RESET. Components may be repositioned as required. B. The Phase B master relay is de-energized when the control switch is placed in RESET. When the master relay is de-energized, components may be repositioned. C. Containment Spray must be reset to allow resetting the master relay for Phase B. Placing the Phase B control switch to RESET will de-energize the slave relays to allow components to be repositioned. D. Initiating condition must clear and control switch must be placed in RESET to de-energize the slave relays that allow components to be repositioned. Proposed Answer: A Explanation (Optional): A. Correct. B. Incorrect. Master relay only operates on the actuation signal. When resetting, slave relays reset C. Incorrect. Spray and CIB have independent resets, even though initiating signal is Page 53 of 200 the same NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Reactor Operator D. Incorrect. If switch is placed in RESET, initiating condition may still exist and the components will still reset Technical Reference(s) OTO-SA-00001 (Attach if not previously ___________ provided) Proposed references to be provided to applicants during None examination: Learning Objective: (As available) Question Source: Bank# Modified Bank # (Note changes or attach _____ parent) New X -----Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge _X ___ _ Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 ---Comments: WTSI 53122 developed but not yet used on NRC exam Page 54 of 200
1. CS&COOL-40302D07011 9
.. -The containment spray system has actuated. The team wants to close MOV-8820A and B, CS pump to spray header isolation valves. Which one of the following is the minimum action the operator must take to ensure MOV-8820A and B remain closed prior to closing the valves? A. Reset the Phase B actuation signal ONLY by depressing a pair of reset push buttons on the MCB. B. Wait one minute since the open signal is only present for one minute after the valves open. C':'" Reset the containment spray actuation signal ONLY by depressing a pair of reset pushbuttons on the MCB. D. Reset both the containment spray actuation signal and the Phase B actuation signal by depressing 2 sets of reset pushbuttons on the MCB. .. ...... --.-.. . Feedback OPS-52102C A. Incorrect -Phase B is a containment isolation signal and does not open any valves including MOV8820AlB B. Incorrect -If the CS actuation signal is not reset then the valves will roll back open C. Correct -Reset the containment spray actuation signal ONLY by depressing a pair of reset pushbuttons on the MeB. P signal or CS actuation opens this valve and remains Sealed in with an R/L logic and is reset with 2 Pushbuttons on the MCB D. Incorrect -this is not the min. since Phase B does not need to be reset A containment isolation phase B and containment spray actuation can be manually initiated from the safeguards panel. There are four PHASE B CTMT ISO CS ACTUATION switches. These switches are grouped in two sets of two switches each. Simultaneously placing both switches in a set in the ACTUATE position will cause a manual containment phase B isolation and containment spray actuation signal. There are four reset push buttons on the safeguards panel--two for containment spray (train A and train B) and two for containment isolation phase B (train A and B). All four push buttons must be momentarily depressed to clear the containment spray actuation and containment isolation phase B signals. Page: 1 of2 9/1812009 Notes 2005 NRC exam 026A4.05 Ability to manually operate and/or monitor in the control room: Containment spray reset switches List the automatic actions associated with the Containment Spray and Cooling System components and equipment during normal and abnormal operations including (OPS40302D07):
Normal control methods

Automatic actuation including setpoint (example SI, Phase-B, LOSP) and the effect of selecting the containment cooler control to local. Page: 2of2 9/18/2009
58. 103K3.03 001/NEW/RO/C/A 3.7/4.1/N/N/3/CVR/Y Refueling on Unit 2 is in progress, and the following conditions exist:
The Containment Equipment Hatch is open and hoses and electrical cables are routed through the Hatch.

Both Main Personnel Airlock doors are open.

The Inner Airlock door is inoperable and cannot be closed.

CTMT Main Purge is in operation.

R-24A and R-24B are discovered to be inoperable.
Which one of the following describes whether or not Fuel movement may continue in Containment, and the reason? A'! CORE ALTERATIONS must be stopped immediately, because R-24A and R-24B are inoperable. B. CORE ALTERATIONS must be stopped immediately, because the Inner Airlock door cannot be closed. C. CORE ALTERATIONS must be stopped immediately, because the Containment Equipment Hatch is open. D. CORE ALTERATIONS may continue in the current condition, because all penetrations are capable of being isolated with manual actions. Page: 154 of 2n 12/14/2009 A-Correct. Either the dampers must be closed or the auto isolation feature must be operable to allow CORE ALTERATIONS. R-24A & R-24B provide automatic isolation on high radiation sensed in the CTMT purge system. B -Incorrect. Only one of the two airlock doors must be capable of closing. Plausible, since at least one airlock door must be capable of closing, but in this case, the outer door is still capable of being closed. C -Incorrect. The eqUipment hatch may be open as long as it is capable of being closed and held in place with 4 bolts in two hours. The hoses and cables give added plausiblity to this distractor, since they would prolong the time that it would take to close the hatch, but they are allowed to be routed through the hatch as long as they have quick disconnects, isolation valves, and blue ownership tags to facilitate closing the hatch in two hours or less if needed for containment closure per STP-18.4, Containment Mid-Loop And/Or Refueling Integrity Verification And Containment Closure, Step 5.2.4 Version 33 and UOP-4.1, Controlling Procedure For Refueling, version 51. D -Incorrect. Plausible, since it is correct for two of the three penetrations listed (Equipment hatch and Personnel Hatch), but the Main purge has an additional requirement of automatic isolation capability per TS 3.9.3, Containment Penetrations, during refueling. R-24A & R-24B being inoperable defeats the required automatic isolation feature. Until the Purge dampers are closed, core alts must stop. Also plausible, since if the dampers were manually closed, the fuel movement could continue, but the choice states "in the current condition" which includes Main purge in operation per the stem. Fuel movement cannot continue with Main purge in operation, but could the dampers were manually closed per TS 3.3.6, Condition C. TSs 3.3.6 Containment Purge and Exhaust Isolation Instrumentation, Amendment No. 146 (Unit 1) & Amendment No. 137 (Unit 2) 3.9.3 Containment Penetrations, Amendment No. 178 (Unit 1) & Amendment No. 171 (Unit 2) Page: 155 c:l277 12/14/2009 Previous NRC exam history if any: 103K303 103 Contai nment System K3 Knowledged the effect that a IO$(J' malfunction d the containment system will have on the following: (CFR: 41.7/45.6) K3.03 Loss of containrnal1: integrity unda-refueling opa-ations. .............. 3.7 4.1 Match justification: This question provides conditions which must be recognized as either a loss of Refueling integrity or allowed for refueling integrity. There is some normal conditions allowed by refueling integrity (but not by Containment integrity in modes 1-4) and one condition that must be recognized as a loss of refueling integrity in Mode 6. Knowledge is required that an automatic iso ckt must be operable for ctmt purge valves OR they must be closed for refueling integrity. The effect of the conditions which must be known to answer this question is that Core alterations are prohibited. Objective: 1 RECALL AND APPLY the LCO and APPLICABILITY for Technical Specifications (TS) or TRM ra:Juirements, and the REQUIRED ACTIONS for 1 HR or lessTS or TRM ra:Juirements, and the relewant portions of BASES that DEFINE the OPERABILITY and APPLICABILITY of the LCO associatoo with the Containment Structure and Isolation System components and attendant a:Juipment alignment, to include the following (OPS-52102A01): Page: 156 d 277
1.6 3.6.1 3.6.2 3.6.3 Contai nment Integrity
-Defi nition Contai nment Containment Air Locks Containment Isolation Valve:, 3.6.4 Contai nment Pressure 3.6.5 Containment Air Temperature 13.6.1 Containment Ventilation System leakage Rate
13.8.1 Contai nment Penetrati on Conductor Overcurrent Protecti ve Dewi Cf!3 (U ni t 2 Only). 12/14/2009 Question # 58 KIA 103K3.03 REFERENCE Docs 3.3 INSTRUMENTATION Containment Purge and Exhaust Isolation Instrumentation 3.3.6 3.3.6 Containment Purge and Exhaust Isolation Instrumentation LCO 3.3.6 The Containment Purge and Exhaust Isolation instrumentation for each Function in Table 3.3.6-1 shall be OPERABLE.
APPLICABILITY: According to Table 3.3.6-1. ACTIONS ----------------------------------------------------------NOTE----------------------------------------------------------- Separate Condition entry is allowed for each Function. CONDITION A. One Required radiation monitoring channel inoperable. B. ------------ NOT E------------ Only applicable in MODE 1, 2, 3, or 4. -------------------------------- One or more Functions with one or more manual or automatic actuation trains inoperable. OR Required Action and associated Completion Time of Condition A not met. Farley Units 1 and 2 A.1 B.1 REQUIRED ACTION COMPLETION TIME Restore the affected 4 hours channel to OPERABLE status. Enter applicable Immediately Conditions and Required Actions of LCO 3.6.3, "Containment Isolation Valves," for containment purge and exhaust isolation valves made inoperable by isolation instrumentation. 3.3.6-1 Amendment No. 146 (Unit 1) Amendment No. 137 (Unit 2) ACTIONS CONDITION C. ------------NOTE------------ C.1 Only applicable during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment. One or more manual channel(s) inoperable. QR Two radiation monitoring channels inoperable. QR Required Action and associated Completion Time for Condition A not met. Farley Units 1 and 2 C.2 Containment Purge and Exhaust Isolation Instrumentation 3.3.6 REQUIRED ACTION Place and maintain containment purge and exhaust valves in closed position. Enter applicable Conditions and Required Actions of LCO 3.9.3, "Containment Penetrations," for containment purge and exhaust isolation valves made inoperable by isolation instrumentation. COMPLETION TIME Immediately Immediately 3.3.6-2 Amendment No. 146 (Unit 1) Amendment No. 137 (Unit 2) Containment Purge and Exhaust Isolation Instrumentation 3.3.6 SURVEILLANCE REQUIREMENTS

NOT E --------------------------------------------------------- Refer to Table 3.3.6-1 to determine which SRs apply for each Containment Purge and Exhaust Isolation Function. SURVEILLANCE FREQUENCY SR 3.3.6.1 Perform CHANNEL CHECK. 12 hours SR 3.3.6.2 Perform ACTUATION LOGIC TEST. 92 days on a STAGGERED TEST BASIS SR 3.3.6.3 Perform MASTER RELAY TEST. 92 days on a STAGGERED TEST BASIS SR 3.3.6.4 Perform COT. 92 days SR 3.3.6.5 Perform SLAVE RELAY TEST. 18 months SR 3.3.6.6 -------------------------------- NOT E ------------------------------ Verification of setpoint is not required.

Perform T ADOT. 18 months SR 3.3.6.7 Perform CHANNEL CALIBRATION. 18 months SR 3.3.6.8 Verify ESF RESPONSE TIME within limit. 18 months on a STAGGERED TEST BASIS Farley Units 1 and 2 3.3.6-3 Amendment No. 180 (Unit 1) Amendment No. 173 (Unit 2) Containment Purge and Exhaust Isolation Instrumentation 3.3.6 Table 3.3.6-1 (page 1 of 1) Containment Purge and Exhaust Isolation Instrumentation FUNCTION 1. Manual Initiation
2. Automatic Actuation Logic and Actuation Relays 3. Containment Radiation Gaseous (R-24A, B) APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS 1,2,3,4, (a), (b) 1,2,3,4 1,2,3,4 (a), (b) REQUIRED CHANNELS 2 2 trains 1 2 SURVEILLANCE REQUIREMENTS SR 3.3.6.6 SR 3.3.6.2 SR 3.3.6.3 SR 3.3.6.5 SR 3.3.6.8 SR 3.3.6.1 SR 3.3.6.4 SR 3.3.6.7 TRIP SETPOINT NA NA -2 s; 2.27 X 10 IlCi/cc (c) (d) -3 s; 4.54 X 10 IlCi/cc (c) (e) -3 s; 2.27 X 10 IlCi/cc (c)(f) 4. Containment Isolation
-Phase A Refer to LCO 3.3.2, "ESFAS Instrumentation," Function 3.a., for all initiation functions and requirements. (a) During CORE ALTERATIONS. (b) During movement of irradiated fuel assemblies within containment. (c) Above background with no flow. (d) With mini-purge in operation. (e) With slow speed main purge in operation. (f) With fast speed main purge in operation. Farley Units 1 and 2 3.3.6-4 Amendment No. 146 (Unit 1) Amendment No. 137 (Unit 2) Containment Penetrations 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Containment Penetrations LCO 3.9.3 The containment penetrations shall be in the following status: Choice C a. Choice B b. Choice D The equipment hatch is capable of being closed and held in place by four bolts; One door in each air lock is capable of being closed; and Each penetration providing direct access from the containment atmosphere to the outside atmosphere either: 1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or 2. capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System. APPLICABILITY: During CORE ALTERATIONS, During movement of irradiated fuel assemblies within containment. Choice A ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more containment A.1 Suspend CORE ALTERATIONS. Immediately penetrations not in required status. Farley Units 1 and 2 A.2 Suspend movement of Immediately irradiated fuel assemblies within containment. 3.9.3-1 Amendment No. 178 (Unit 1) Amendment No. 171 (Unit 2) Containment Penetrations 3.9.3 SURVEILLANCE REQUIREMENTS SR 3.9.3.1 SR 3.9.3.2 SR 3.9.3.3 SURVEILLANCE FREQUENCY Verify each required containment penetration is in the 7 days required status. Verify each required containment purge and exhaust 18 months valve actuates to the isolation position on an actual or simulated actuation signal. --------------------------------N 0 TE----------------------------- 7 days Only required for an open equipment hatch. Verify the capability to install the equipment hatch. Farley Units 1 and 2 3.9.3-2 Amendment No. 165 (Unit 1) Amendment No. 157 (Unit 2)
59. G2.1.17 001/NEW/RO/M EM 3.9/4.0/N/N/2/CVR/Y Which one of the following demonstrates proper communication techniques that ensure information is transmitted and received effectively lAW ACP-1.0, Plant Communications?
A. The UO speaking to an extra SO: "Go check out the alpha charging pump and let me know if it is ready to start". SO: "Check out the alpha charging pump and let you know when it's ready to start". UO: "That's correct". The UO speaking to the Rover: "Check one bravo CCW heat exchanger outlet isolation valve, 01 P17V008 Bravo, closed". Rover: "Check one bravo CCW heat exchanger outlet isolation valve, 01 P17V008 Bravo, closed". UO: "That's correct". C. The SS speaking to both Control Room Operators: "Secure the one charlie reactor coolant pump". OATC: "Secure the one charlie reactor coolant pump". SS: "That's correct". The UO obtains a peer check and secures the one charlie reactor coolant pump. D. The SS speaking to the OATC: "Secure containment minipurge supply and exhaust fans". OATC secures containment minipurge supply and exhaust fans, and then reports to the SS: "Containment minipurge supply and exhaust fans are secured". SS: "Containment minipurge supply and exhaust fans are secured". OATC: "That's correct". A -Incorrect. The unit number is required by step 5.1.3 of ACP-1.0 to ensure the proper component is operated. This is especially important since an extra operator is being addressed, but would be required even for the Rad Side System Operator. Plausible, because the UO is assigned to a specific unit, and it may seem that it is obvious which unit is intended, but a unit is required to prevent wrong unit component manipulation events regardless of who is speaking or who is being addressed. B -Correct. The phonetics are normally required for all letters, except for in a TPNS and some other exceptions listed in STEP 5.1.3 & & APPENDIX 1 of ACP-1.0. The example given is correct in that it does not require a directed communication since no one else is present, it has a unit deSignator for the "1 "B CCW valve (and a "1" in the TPNS), and the B on the end of the TPNS has a phonetiC pronunciation (Bravo). Page: 157 of 277 12/14/2009 That is what is required by ACP-1.0. C -Incorrect. This communication must be "directed" using the person's name or title to ensure the appropriate person receives the instruction per step 5.1.4 "When more than two people involved in the same task are in the immediate area then use the name or title of the intended receiver prior to the message or instruction." Plausible, since normally eye contact is made while speaking and is sufficient in other than instructional communication. It is a common misconception that if both people think they know who is being spoken to they don't need to direct the communication. However, whenever more than one person is in the area, to ensure information is transmitted and received effectively, ACP-1.0, step 5.1.4 requires a name or title to precede this communication. Also, the operator (OATC) who repeated the communication back must perform the action, or use another 3 way communication to the UO prior to the UO completing the action. D -Incorrect. A repeat back must be obtained prior to the action being performed for instructional communications per step 5.1.5 & 5.1.6. Plausible, since no repeat back is obtained after informational communications per step 5.1.7. This sequence described is acceptable in the case of soliciting a meter reading when no action is required other than obtaining the reading and communicating the value; SS asks for a meter reading, prior to repeating back, the operator obtains and communicates the meter reading, SS acknowledges the meter reading, operator states: "that's correct". ACP-1.0, Versioo 5.0 5.1 The foil owi ng communi cati on techni ques are provi ded to ensure i nformati on is transmitted and received effectively. 5.1.3 Be spE£ific when identifying Equipment. Identify Equipment by using noun name, TPNS designation, or approved abbreviation such as RWSf for the refueling water storage tank. Approved abbreviations are found in FNP-O-AP-25, EQUIPI'v1ENT IDENTIFICATION. Specify if it isUnit 1, Unit 2 a g,ared equipment. (NOTE: If the unit designation is included in the Equi pment name, i.e. 2A chargi ng pump, that is ccceptabi e.) Bei ng spoci fi c ensures that the information is detailed enough 00 that the correct component is identified. When a loog string of alpha-numeric char ader s is being spoken SJch asa TPNS number it is acceptableto ooly use the phooetic alphabet fa the last letter in the sequence. See Appendix 1 examples for clarification. 5.1.4 Ensure the intended individual receives the message. When mae than two people involved in the same task are in the immediate area then use the name a title of the intended receiver prior to the message or instruction. 5.1.6 As the receiver of an instructional communication, the individual should CEknowledge receipt of the communication by providing feedbCEk in the form of repeatbCEk or paraphrcse. V erbati m repeatbCEks are not rEqui red, but may be useful in oome <Eti vi ti es ACP-1.0, APPENDIX 1, PageS of 5, VersiooS.O These are examples of how the phonetic alaphabet should be used: Stating a TPNS designatioo: Page: 158 d 2n 12/14/2009 Written: 01 E21VOO9C Spoken: 01 E21VOO9 Charlie Previous NRC exam history if any: G2.1.17 G2.1.17 Ability to make accurate, dear, and conciS! verbal reports (CFR: 41.10/45.12/ 45.13) RO 3.9 SRO 4.0 Match justification: This question requires knowledge of the techniques which the applicant is required to utilize in verbal communications which ensure accurate, clear, and concise verbal reports. ACP-1.0, an FNP Administrative Control Procedure, has a section titled: 5.1 "The following communication techniques are provided to ensure information is transmitted and received effectively". This question requires knowledge of these ACP listed techniques. Objective:
1. Expl ai n the i mportarlC9 of mai ntai ni ng professi onal communi cal:i ons when usi ng pi arlt communical:i ons pment (0PS40502C01).

2. Outl i ne management's expectati ons for communi cal:i ons (0PS40502C02).

3. Expl ai n the purpose of arld the method for conducti ng three-way communi cal:i ons (0PS40502C03) . Page: 159 d 277 12/14/2009 Question # 59 KIA G2.1.17 REFERENCE Docs 06/26/06 16:55:45 FNP-O-ACP-I.O 5.0 General Communications Guidelines 5.1 This section addresses verbal communications that are instructional in nature. All Distractor A Distractor C
plant personnel are encouraged to use these communication techniques at times other than required by this procedure so that these techniques become a regular habit. Three-way communication is the expected standard for communications involving directions, operations or transmission of technical data. Refer to Appendix 1 of this procedure for examples of effective Three-way communication techniques.
The following communication techniques are provided to ensure information is transmitted and received effectively. 5.1.1 5.1.2 5.1.3 5.1.4 Make verbal communications clear, complete and concise. Table I shows a suggested phonetic alphabet and is provided as an aid to enhance clear and concise communications. Keep verbal instructions limited in scope. Break complicated instructions or mUltiple actions into simple steps containing single actions or communicate them in writing. Be specific when identifying equipment. Identify equipment by using noun name, TPNS designation, or approved abbreviation such as RWST for the refueling water storage tank. Approved abbreviations are found in FNP-0-AP-25, EQUIPMENT IDENTIFICATION. Specify if it is Unit 1, Unit 2 or shared equipment. (NOTE: If the unit designation is included in the equipment name, i.e. 2A charging pump, that is acceptable.) Being specific ensures that the information is detailed enough so that the correct component is identified. When a long string of alpha-numeric characters is being spoken such as a TPNS number it is acceptable to only use the phonetic alphabet for the last letter in the sequence. See Appendix I examples for clarification. Ensure the intended individual receives the message. When more than two people involved in the same task are in the immediate area then use the name or title of the intended receiver prior to the message or instruction. 5.1.5 The sender of information is responsible for ensuring that the receiver of information gives a paraphrased repeat back. The sender is then to acknowledge the receiver by verifying the information is correct. This Version 5.0 06/26/06 16:55 :45 FNP-O-ACP-l.O will result in the "three-way" communication loop being completed as expected. Version 5.0 06126/06 16:55:45 FNP-0-ACP-l.0 D distractor plausibilty 5.1.6 5.1.7 5.1.8 5.1.9 5.1.10 As the receiver of an instructional communication, the individual should acknowledge receipt of the communication by providing feedback in the form of repeatback or paraphrase. Verbatim repeatbacks are not required, but may be useful in some activities. The intent of this step is not to delay "Immediate Actions" until a repeat back has been made. Immediate Action steps should be performed when required, and then communicated to the appropriate individual. Informational communication normally does not require acknowledgement. However, it is appropriate for the sender of information to solicit acknowledgement when it is deemed necessary. The use of sign language is discouraged except in accepted industry practices such as crane operations, or in extremely high noise areas. Individuals using sign language should ensure understanding of the signs to be used prior to commencing the evolution. Do not assume that other members of your team understand what you are doing or thinking. Proper communications among team members ensures that each person knows what the others are doing. The use of profanity is not conducive to enhancing the quality of communication and is offensive to many people. Profanity degrades the professional atmosphere and therefore shall not be used. 5.2 Plant Public Address System (Gaitronics) 5.2.1 Minimize its use as the paging system to reduce background noise. 5.2.2 Channel 5 is designated for emergency use only. 5.2.3 In situations where Operations personnel deem it necessary that an immediate response be obtained, it is appropriate to announce over the public address system for an individual to pickUp on line 5. This line's designation as emergency use only should imply to the person being paged that immediate response is expected. 5.2.4 The Public Address System may be used to inform or update plant personnel of the status of an abnormal or emergency condition, change in plant status, or a major plant event in progress or anticipated. Version 5.0 06/26/06 16:55:45 APPENDIX 1 These are examples of how the phonetic alaphabet should be used: Stating a TPNS designation: Written: Spoken: QIE21V009C QIE21V009 Charlie Stating a breaker designation: Written: EA08 Spoken: Echo Alpha 08 Stating a 7300 card designation: Written: Spoken: C-8 145 Charlie 8 145 Stating a conduit designation: Written: Spoken: NHS 940 NHS 940 Stating a work order status Written: Spoken: PI Papa I Page 5 of 5 B, correct choice FNP-O-ACP-I.O Version 5.0
60. G2.1.45 001/NEW/RO/MEM 4.3/4.3/N/N/4/HBF/Y The following plant conditions exist on Unit 1: AT 1000:
N-41 , N-42, N-43, and N-44, PR Nuclear Power, indicate 100% power.

Main Generator Load is 901 MW.

All SG steam flows are 4.1 X 10 6 Ibs/hr. AT 1010: The crew identifies PK-3371A, 1A SG Atmospheric Relief Valve Controller, failed to 100% demand.

The UO places PK-3371A in manual and lowers the demand to 0%.

The crew suspects the Atmospheric Relief Valve has not closed. Which one of the following sets of stable plant parameters indicates that PCV-3371A has remained OPEN? Stm Flow (X 10 6 Ibs/hr) 1ASG 1B SG 1C SG MW FI-474 FI-484 FI-494 A. 840MW 4.970 4.100 4.100 B. 840MW 4.404 4.398 4.398 C. 880MW 4.500 4.100 4.100 D'!" 880MW 4.254 4.248 4.248 Page: 160 c:I 277 12/14/2009 Knowledge:
an Atmospheric Relief has a design capacity of 3% total steam flow. Distractor uses Safety valve design capacity of 7.6% total steam flow. Stm flow indication: 1A SG will demonstrate a minutely higher steam flow than the other 2 SGs only because of the head loss which occurs in the Cross-over piping (42" line in MSVR). INITIAL Steam flow @100% =12.3 Mlbh]/ 3) =4.1 X 10 3 lbs/hr
3% steam flow = 12.3 Mlbh/100%*3%=
.369 Mlbh
7% steam flow = 12.3 Mlbh/100%*7%=
.861 Mlbh Fundamental: Pascal's LAW states that the pressure exerted within a system is felt equally and undeminished throughout that system. Pascal's Law is designated for a closed system, but the fundamental is relatively consistent in an open system when conservation of Mass Flow is maintained. A -Incorrect. This is the MW loading expected after a steam leak of 7%. An Atmospheric relief capacity is only 3%. Additionally, since the MSIVs are open, the steam flow will be balanced between the three SGs, with only a minor difference in A SG steam flow due to head loss. ---! pszgl = . 10 195.16%! = 857 MW ( 7% ]'750 ' ' 484°/ [901MW] 1 o 85psig resulting in a MW load of: 100010 ADDED a 20 MW loss to increase desparity between answers. B -Incorrect. This is the MW loading and steam flows expected after a steam leak of -7%. C -Incorrect. Although the MW load is correct, the steam flow would be shared by all SGs since cross-connected. D -Correct. the design capacity of an Atmospheric relief valve is 3% total steam flow at 1035 psig; Since 100% power Steam pressure is 750 psig, then the Impact from a failed open ARV will be ---I. pszg,l=. 10 197.73%1 = 881MW ( 3% ]'750 . . 2 170/ [901MW] 1035psig resulting in MW reduction of: 100% Validated on Laptop Sim 9/23/09 with IC-73 --actual MW load is 880 MW and steam flow 1 A == 4.2, 1 B=4.2, 1 C==4.1 (smoothed homepage view) instrument view has swing variance but avg around 4.25 on all instrumentation) Page: 161 cl2n The Steam flow will be shared by all Steam generators: With only a minor variation noted in the 1 A SG steam flow noticable. 12/14/2009 Previous NRC exam history if any: N/A G2.1.45 2.1.45 Ability to identify and interpret diverse indicationsto validate the response of another indication. (CFR: 41.7/43.5/ 45.4) RO 4.3 SRO 4.3 Match justification:
PC-3371 AlBIC from the MCB requires evaluation of diverse indications to validate success since there are no Position indicators on the MCB. the indication that must be validated with diverse indications is the MA station 0% demand.

The indications that must be interpreted are Steam flow and turbine load. Objective:
OPS-40201 A02; relate and identify the operational characteristics including design features, capacities, and protective interlocks for the components associated with the Main and Reheat Steam System [ ... ]. OPS-52521 007; Analyze plant conditions and Determine the successful completion of any step in AOP-14 [ ... J. Page: 162 of 2n 12/14/2009 Question # 60 KJA G2.1.45 REFERENCE Docs FNP-FSAR-10 outside the containment. These valves are Safety Class 2A and Category I Seismic. The range of pressure settings of the safety valves on each line is in equal increments from 1075 psig to 1129 psig. The maximum actual capacity of a single valve fully open at 1085 psig is 890,000 Ib/h. 10.3.2.2.4 Relief Valves Installed on each main steam line upstream of the main steam isolation valves and downstream of the safety valve is one atmospheric relief valve. The valves are pneumatically actuated and sized to pass 405,500 Ib/h of steam at 1025 psig. They are capable of going from fully closed to fully open within 35 seconds or less. The valves are also capable of being modulated over the pressure range of 1085 psig to 100 psig. Valve control is automatic by steam line pressure with remote manual control of the setpoint. A local manual operator is provided for valve operation in the event of complete loss of automatic control. An emergency source of control air is provided to enable remote manual operation. 10.3.3 EVALUATION Following a sudden load rejection of up to 50 percent, the MSSS prevents a reactor trip by bypassing the steam directly to the condenser as described in subsection 10.4.4. Following a turbine trip or load rejection above 50 percent or when the turbine bypass system is not available, the MSSS effects a safe reactor trip by removing excessive heat from the reactor coolant through the exhausting of secondary steam through atmospheric power-operated relief valves and the spring-loaded safety valves. The power-operated relief valves and the spring-loaded safety valves also protect the steam generator and the main steam piping from overpressure. In the unlikely event of a main steam line rupture, the isolation valves in the main steam lines provide steam line isolation, as described in subsection 5.5.5. The valving safety requirements are established to cover the following situations: A. Break in the Steam Line From One Steam Generator Inside Containment The steam generator associated with the damaged line will discharge completely into the containment. Without reverse flow protection, the other steam 10.3-6 REV 15 FNP-FSAR-10 downstream point in the steam line before the containment. C. The portion of the MSSS up to and including the main steam isolation valves is necessary for the safe shutdown of the plant and is Safety Class 2A and Category I Seismic. D. Uncontrolled steam release as a result of a steam line failure is limited to the contents of one steam generator in order to keep the related effect upon the reactor core within prescribed bounds. E. The failure of any main steam line or malfunction of a valve installed therein will not: 1. Render inoperable any engineered safety feature (ESF) . 2. Result in the containment pressure exceeding the design value or impairing its integrity. Other safety-related design provisions include: A. The steam generator safety valves. B. The steam generator relief valves. C. The steam supply to the turbine-driven auxiliary feedwater pump. The steam supply to this turbine has a safety classification because of the safety-related functions of the auxiliary feedwater system. The turbine steam supply lines are connected to two steam generator steam lines upstream of the steam line protective valving to provide both redundancy and dependability of supply. Isolation valves in each line maintain the separation of the main steam lines by preventing any interconnecting backflow. D. Each steam generator includes an internal restriction which acts to limit the maximum flow and the resulting thrust forces created by a main steam line break. 10.3.1.3 Design Data A. Number of steam generators 3 B. Total flow (lb/h) 12.26 E+06 10.3-3 REV 15 FNP-FSAR-10 Each valve fails closed on loss of electrical or air supply. They are capable of modulating over the pressure range of 100 to 1085 psig with a stroke time of 35 seconds or less. Each valve is sized to pass a total of approximately 405,500 Ib/h of steam (10 percent of plant maximum calculated steam flow) at the no-load pressure of 1025 psig. Additionally, the maximum actual capacity of any single valve at an inlet steam pressure corresponding to the steam generator shell design pressure (1085 not exceed 890, Ib/h to Ilml t the steam release If anyone vciil'ctilafedfiow ....... ". .' , .... ' open. ./ '.' .... d':' Discharge from the power-operated .=.4i086'Mlbm/hr'ISG . s in addition to the normal operating con :.:,; . d . . h ... Durlng emergency con ltlons, t e maln s ............... ' *..............
,. relief valves provide a means to contro generator pressures, and, or cool down ...........................
1:"" ............. ----...... '"""f ...... --..... ---.1I plant conditions these relief valves also give the plant flexibility of operation and the capability for a controlled cooldown. Isolation valves are provided upstream of each valve to allow maintenance. During a period when all other valves are out of service, the steam generator safety valves provide the necessary relieving capability. Prior to shipment, each valve was hydrostatically tested in the manufacturer's facilities in accordance with the applicable code. Leakage was 40 cm 3/h in the valves that were tested. During plant operation, the valves are accessible for inspection. The operability of the alternate air supply system may be demonstrated during refueling shutdowns by using the alternate air supply to cycle open and closed each of the power-operated relief valves. 10.3.9 MAIN STEAM ISOLATION VALVES The main steam isolation valves consist of two swing-disc check valves in each of the three main steam lines. These valves are located outside of the containment downstream of the main steam safety valves. The main steam line isolation valves and their bypass valves are designed to stop forward flow and to isolate the steam generators and the main steam lines on signal initiated by engineered safety features actuation system under any of the following conditions: 10.3-13 REV 15 07/02/09 07 :25: 19 FNP-2-UOP-3.1 NOTE: When the feedback loop is placed inlout of service Target and Demand are reset to equal Pimp. Thus any RAMP or HOLD in progress will be stopped. As a result the operator will have to reenter the desired Target and Rate. IMP PRESS LOOP should remain in service until >97% power to avoid control issues associated with #4 GV when initiating a ramp from near 95% (CR20071 05465). 5.20 WHEN >97% power AND the IMP PRESS LOOP is in service, THEN remove the feedback loop as follows: NOTE: 5.20.1 5.20.2 5.20.3 5.20.4 Check Turbine on HOLD. On the FEEDBACK STATUS DISPLAY, move the cursor to IMP PRESS LOOP IN. Depress the SELECT key and verify IMP PRESS LOOP is highlighted in reverse video. Depress the STOP key and verify the FEEDBACK STATUS indicates IMP PRESS OUT.
During steady state power operations the reactor core power level is limited to 2775 mega watts thermal. (NRC Regulatory Issue Summary 2007-21)

Minor fluctuations of turbine Ire actor power are expected due to grid fluctuations and normal tolerances in control systems. Unplanned increases in reactor power due to positive reactivity addition shall be promptly terminated with turbine load reduction, control rod insertion andlor boration.
Reactor power shall be returned to at or below the previous steady state power level.
DEH demand of 923 MW will cause #4 GV to go to full open position 5.21 Increase power to 100%, ensuring TAVG does not exceed 577.2°F. 5.21.1 WHEN required, THEN take prompt action to compensate for excessive positive reactivity additions.
5.21.2 When available, the 15 minute average core thermal power (QC462I M 15) should be monitored and maintained below 2775 MWth. If this value is exceeded, the OA TC must reduce power to ensure the 1 hour average power remains below the limit of2775 MWth. 5.21.3 In no case should 102% power be exceeded. Version 86.0 07/02/09 07 :25: 19 FNP-2-UOP-3.1 3.2 Turbine 3.2.1 Do not exceed the turbine generator loading rates specified by FNP-2-S0P-28.1, TURBINE GENERATOR OPERATION. 3.2.2 The following apply for condenser pressure conditions:
WHEN the turbine is operating at:::: 30% load, THEN the maximum permissible condenser pressure is 5.5 inches Hg. (2.7 psia).

WHEN the turbine is operating at < 30% load, THEN the maximum permissible condenser pressure is 3.5 inches Hg. (1.7 psia).

Refer to FNP-2-AOP-8.0, PARTIAL LOSS OF CONDENSER VACUUM, for remedial actions (Westinghouse Customer Advisory Letter 86-02.) 3.2.3 The following DEH valve position limits apply

The DEH valve position limit shall be maintained at 115% while ramping or when at 100%.

The DEH valve position limit should be maintained 8 to 10% above valve position demand when holding for more than 24 hours at some power level less than 100%. 3.2.4 The DEH system queries the valve LIMIT LOWER and RAISE keys in 1 second intervals.
Therefore when adjusting the valve position limit a wait seconds should be allowed after the key is released before depressing the key again to ensure the DEH system recognizes the key had been released. {AI2002200059}. 3.2.5 Operate the MSR controls per FNP-2-S0P-28.1, TURBINE GENERATOR OPERATION, Precautions & Limitations. 3.2.6 When the Impulse Pressure feedback loop is placed in/out of service Target and Demand are reset to equal Pimp. Thus any RAMP or HOLD in progress will be stopped. As a result the operator will have to reenter the desired Target and Rate. IMP PRESS LOOP should remain in service until >97% power to avoid control issues associated with #4 GV when initiating a ramp from near 95% (CR2007105465) Version 86.0 FNP-FSAR-10 A steam generator partition factor of 0.1 and a condenser air ejector partition factor of 10-4 have been used in the evaluation of environmental consequences of postulated accidents (e.g., Steam Generator Tube Rupture, subsection 15.4.3.4). These are conservative values for the range of water chemistry allowed by WCAP-7452 based on measurements made at operating Westinghouse plants. 10.3.6 INSTRUMENTATION APPLICATIONS The steam flow restrictors installed in the steam generators are also used for steam flow measurements during normal operation. Two flow transmitters and two pressure transmitters are installed in the main steam line from each steam generator. The steam flow and pressure signals are fed into reactor protection and feedwater control system circuits to control the feedwater flow to each steam generator, to close the isolation valves in case of rupture in main steam lines, and to open the power-operated relief valves in case of overpressure. 10.3.7 MAIN STEAM SAFETY VALVES Overpressure protection for the three steam generators is provided by the main steam safety valves. The design basis for the main steam safety valves is that they must have sufficient capacity so that main steam pressure does not exceed 110% of the steam generator shell-side design pressure. Based on this requirement, the valves are sized to relieve 105% of the maximum calculated steam flow at an accumulation pressure not exceeding 110% of the steam generator shell design pressure. Design parameters for the main steam safety valves are given in table 10.3-1. Due to the large mass flowrate, each steam generator is protected by a number of valves. The maximum actual capacity of a single valve fully open at 1085 psi gauge does not exceed 890,000 Ib/h. This provision serves to limit steam release if anyone valve inadvertently sticks open. The main steam safety valves are located on the main steam lines outside the containment and upstream of the main steam isolation valves. Each of the three main steam lines is equipped with five safety valves. To prevent chattering during operation of the safety valves, each of the five valves on a steam line is set at a different set pressure. The first valve set pressure is 1075 psig, which corresponds to the steam generator shell design pressure minus the pressure loss from the steam generator to the valve. Each of the remaining valves is set at a higher pressure such that all valves are open and 10.3-10 REV 15
61. G2.1.9 001/NEW/RO/M EM 2.9/4.5/N/N/2/CVR/Y Which one of the following lists ONLY those personnel in which BOTH MUST request permission to enter the Control Room At-the-Controls Area from the OATC lAW NMP-OS-007-001, Conduct of Operations Standards and Expectations?
A. NRC Inspectors, Plant Manager B. Operations Superintendents, Site VP C. Operations Manager, Chemistry Foreman Reactor Engineer, Health Physics Foreman A -Incorrect. NRC inspector and Plant Manager are incorrect per NMP-OS-007-001, step 6.11.2.1. Plausible, since not all of management or NRC/INPO observers are exempt from getting permission, but the resident NRC inspector and the Plant Manager are both exempt. B -Incorrect. Operations Superintendents & Site VP are incorrect per NMP-OS-007-001, step 6.11.2.1. Plausible, since not all of management are exempt from getting permission, but the Operations Superintendents & Site VP are both exempt. C -Incorrect. Operations Manager is incorrect per NMP-OS-007-001, step 6.11.2.1. Chemistry Foreman is correct since the Chemistry Foreman is NOT exempt and must get permission prior to entry. D -Correct. Correct per NMP-OS-007-001, step 6.11.2.1. NMP-OS-007-001, Version 5.0 6.11 .2.1 Access Protocol Personnel who are exempt from requesting permission to enter the ATCA include:
Site VP

Plant Manager

Operations Manager and Operations Superintendents

On-duty shift operating crew, including the ST A

NRC Inspectors All others wishing to enter the ATCA must obtain permission from a licensed operator on shift Page: 163 c:I 2n 12/14/2009 Previous NRC exam history if any: G2.1.9 2.1.9 Ability to direct perSlrlnei adivities inside the control room. (CFR: 41.10/45.5/
45.12/ 45.13) RO 2.9* SAO 4.5 Match justification: To answer this question correctly, knowledge is required of who needs to obtain permission to enter the control room At the Controls Area and who does not. The control room staff must know this to control access to minimize distractions and properly direct activities inside the control room per NMP-OS-007-001 Step 6.11.1: "Access to the main control room is managed so operators are not distracted from properly monitoring plant parameters." Objective:
6. Describe Management'sexpectationsfor Control Room Formality (0PS40502C06).

7. Describe the "at the controls area," and explain the controls associated with axessing this area (0PS40502C07).
Page: 164 d 2n 12/14/2009 Question # 61 KIA G2.1.9 REFERENCE Docs Southern Nuclear Operating Company SOUTHERN A. Nuclear Conduct of Operations NMP-OS-007-001 Management Version 4.0 Standards and Expectations COMPANY Energy IOStrvt Your \VorlJ" Instruction Page 17 of 45 6.10.2.2 Control Room Supervision The Shift Manager is the senior management representative on shift and has primary responsibility for the safe operation of the facility. The Shift Supervisor and Shift Manager maintain awareness by walking down control room boards on a frequency no less than once per shift. Periodically, each Shift Supervisor should tour the plant areas outside of the Control Room, but should be able to respond to the control room promptly. 6.11 Main Control Room Access 6.11.1 Standard Access to the main control room is managed so operators are not distracted from properly monitoring plant parameters. 6.11.2 Expectations 6.11.2.1 Access Protocol Access to the control room area is limited to personnel requiring access for official business in order to avoid distractions to operators. The duty Shift Supervisor or Shift Manager is authorized to refuse entry, or direct individuals to leave, in order to maintain reactor control and minimize distractions. At the Controls Area (A TCA) and Control Room Boundaries are defined in local procedures. Personnel who are exempt from requesting permission to enter the A TCA include: Distractor A, B &C
Site VP D

Plant Manager

Operations Manager and Operations Superintendents

On-duty shift operating crew, including the ST A

NRC Inspectors All others wishing to enter the A TCA must obtain permission from a licensed operator on shift. Once an individual has obtained permission to enter the A TCA, that person may exit and enter without re-authorization during the shift as long as the purpose has not changed. The on shift control room crew compliment is granted access to the ATCA without permission.
The on-shift Shift Supervisor is responsible for maintaining the activities in the control room at an appropriate level. The Shift Supervisor, in conjunction with the control room staff, continually evaluates plant conditions, ongoing maintenance, testing, and the number of personnel in the control room to ensure the control room crew is able to maintain operational focus.
62. G2.2.3 001/FNP BANK/RO/M EM 3.8/3.9/N/N/2/CVR/Y Both Units are operating at 22% power with the following conditions:
Both units 4160V Busses A, Band C are powered from their respective Startup Transformers.
AT 1000 the following occurs: Due to Severe Weather, the 1 B Startup Transformer and 2A Startup Transformer became de-energized. Which one of the following states ALL of the Reactor Coolant Pumps (RCPs) which will still be running after the event? 1A RCP 2ARCP B. 1A RCP 2B and 2C RCPs C. 1 Band 1C RCPs 2ARCP D. 1 Band 1C RCPs 2B and 2C RCPs Page: 165 of 277 12/14/2009 A -Correct. 2A RCP and 1A RCP. Since 2A and 1 B SU xformer trips this means 2B and 2C and 1 Band 1 C RCPs will be tripped. Therefore the 2A and 1A RCPs will be running. B -Incorrect. Plausible, since this would be correct if the unit 2 SU XFMR power configuration was the same as for Unit 1. C -Incorrect. Plausible, since this would be correct if the unit 1 SU XFMR power configuration was the same as for Unit 2. D -Incorrect. Plausible, since this the RCPs that lose power, not which ones are still running after the others lose power. This is for the 2B SU and the 1 A SU tripping. Unit 2 -2A SU Transformer supplies power to the 2B RCP on the 2B 4160V Bus and the 2C RCP on the 2C 4160v bus. Unit 1 -1 B Startup Transformer supplies power to the 1 B RCP on the 2B 4160v bus and 1 C RCP on the 2C 4160v bus. This is a difference between the units in that U-2 has a different SU xformer supplies to the RCP busses than U-1 . Unit 1: S/U XFMR 1A 18 18 1A 18 Unit 2: S/U XFMR 2B 2A 2A 2A 2B Page: 166 of 277 4160V Bus 1A 18 1C 1D 1E 4160V Bus 2A 28 2C 20 2E 12/14/2009 Previous NRC exam history if any: G2.2.3 2.2.3 (multi-unit license) KnOlNledge of the deSgn, procedural, and operational differences between units (CFR: 41.5/41.6/41.7/41.10/45.12) RO 3.S SRO 3.9 Match justification: This is a difference in power supplies to the RCPs on each unit and tests the knowledge of those differences. Objective: 1 NAM E AN 0 I 0 ENT I FY the Bus power suppl i es, for those eI ectri cal componoots asoociatoo with the Intermooiateand Low VoltageAC Distribution System, to incl ude the following (OPS-40102B04):
4160V AC Buses

600V Load Control Cooters

6OO\480\20SV Motor Control Cooters Page: 167 of 277 12/14/2009 Question # 62 KJA G2.2.3 REFERENCE Docs FNPUNIT 1 LOAD LIST A-S062S0 lA 4160VBUS AB -139' D177002 BKR TPNS DESCRIPTION SEE PAGE N1R15AOOO1-N lA 4160V BUS DAOl N1R12AOS02-N lB UNIT AUX TRANSFORMER (NORMAL) <<< DA02 NIR15BKRDA02 PT COMPARTMENT condition of stem: -DA03 NIN21MOOOIA-N lA CONDENSATE PUMP The RCP buses are DA04 NIB41MOOOIA-N lA REACTOR COOLANT PUMP supplied from the SUT --DA05 NIP26MOOOIA-N lA CIRC WATER PUMP DA06 N1RllBOOO7-N II 4160/600V SST >>> EI02 A-2 DA07 N1RllAOS01-N lA STARTUP TRANSFORMER (ALTERNATE)
<<< 1 B SU AND 2A SU NO POWER. 1A BUS UNAFFECTED-- 1A RCP REMAINS RUNNING. Correct Pt 1 lsecta.doc Page A-I Rev. 2 FNPUNIT 1 LOAD LIST A-506250 IB 4160V BUS AB -139' C177003 BKR TPNS DESCRIPTION Nl.R15AOOO2-N 1B 4160V BUS DB01 N1R12A0502-N 1B UNIT AUX TRANSFORMER (NORMl!\T \ /// condition of stem: OB02 NIR15BKROB02 PT COMPARTMENT The RCP buses are OB03 NIB41MOOOIB-N IB REACTOR COOLANT PUMP supplied from the SUT OB04 NIP26MOOOIB-N IB CIRC WATER PUMP I DB05 N1R11A0502-N 1B START-UP TRANSFORMER (ALTERNATE) <<< 1 B SU AND 2A SU NO POWER. 1 B BUS AFFECTED--1 B RCP no power. C&D distractors lsectb.doc Page B-1 Rev. 2 FNPUNIT 1 LOAD LIST A-506250 lC 4160V BUS AB -139' C177004 BKR TPNS DESCRIPTION SEE --PAGE N1R15AOOO3-N lC 4160V BUS DC01 N1R11A0502-N 1B START-UP TRANSFORMER (NORMAL) <<< DC02 NIRI5BKRDC02 PT COMPARTMENT DC03 NIB4IMOOOIC-N IC REACTOR COOLANT PUMP DC04 N1R12A0502-N 1B UNIT AUX TRANSFORMER (ALTERNATE) <<< DC05 NSR31G0501-N 10 4160/600V SST >>> EOO2 C-2 N1R11G0510-N 1T 4160/600V SST >>> ET02 C-50 .... I C&D distractors lsectc.doc Page C -1 Rev. 3 FNPUNIT2 LOAD LIST A-351199 2A4160VBUS AB -139' D-207002 BKR TPNS DESCRIPTION SEE PAGE N2R1SAOOO1-N 2A 4160V BUS DA01 N2R12AOS02-N 2B UNIT AUX TRANSFORMER <<< DA02 N2R15BKRDA02 PT COMPARTMENT DA03 N2N21MOOO1A-N 2A CONDENSATE PUMP DA04 N2B41MOOO1A-N 2A REACTOR COOLANT PUMP DA05 N2P26MOOO1A-N 2A CIRC WATER PUMP DA06 N2R11BOOO7-N 2I 4160/600V SST >>> EI02 (NORMAL) A-2 DA07 N2R11AOS02-N 2B STARTUP TRANSFORMER <<< A&C choices u2mastra.doc Page A-I Rev. 2 FNP UNIT 2 LOAD LIST A-351199 2B 4160VBUS AB -139' D-207003 BKR TPNS DESCRIPTION N2R15AOOO2-N 2B 4160V BUS DBOl N2Rl2AOS02-N 2B UNIT AUX TRANSFORMER <<< DB02 N2R15BKRDB02 PT COMPARTMENT DB03 N2B41MOOOIB-N 2B REACTOR COOLANT PUMP DB04 N2P26MOOOIB-N 2B CIRC WATER PUMP DBOS N2RllAOSOl-N 2A START-UP TRANSFORMER <<< B&D choices u2mastrb.doc Page B-1 Rev. 2 FNP UNIT 2 LOAD LIST A-351199 2C 4160VBUS AB -139' D-207004 BKR TPNS DESCRIPTION SEE PAGE N2a15AOOO3-N 2C 4160V BUS DCOl N2RllA050l-N 2A START-UP TRANSFORMER <<< DC02 N2R15BKRDC02 PT COMPARTMENT DC03 N2B41MOOO1C-N 2C REACTOR COOLANT PUMP DC04 N2R12AOS02-N 2B UNIT AUX TRANSFORMER <<< DCOS NSR1.1BOS2S-N LOW LEVEL RADWASTE STORAGE FACILITY C-2 4l60-120/20BV XF.MR >>> FUSED DISC SW NSR1BSOS46-N >>> lTT AC DIST CAB >>> B&D choices u2sec c.doc Page C-l Rev. 2
63. G2.2.36 001/N8N/RO/C/A 3.1/4.2/N/N/3/CVR/Y Unit 1 is in Mode 6 with refueling in progress, and the following conditions occurred:
At 1000:
The 1 B DG is tagged out for Maintenance.

The 1A RHR pump is in operation.

The 1 B RHR pump is in standby. At 1005:

DG15, 1 B SU XFMR to 1 G 4160V BUS, tripped open. Which one of the following correctly states whether or not the Tech Specs listed below are met?

3.8.2 AC Sources-Shutdown

3.9.4 Residual Heat Removal (RHR) and Coolant Circulation-High Water Level TS 3.8.2 TS 3.9.4 A'! is met is met B. is met is NOT met C. is NOT met is met D. is NOT met is NOT met A -Correct. Only one offsite transmission line is required in Mode 6 during refueling, so with the 1 A SU XFMR still operable, the TS 3.8.2 is met. Only one train of RHR is required to be operable and in operation at this refueling water level, so with the 1 A RHR still operating and in operation, the TS 3.9.4 is met. B -Incorrect.
The first part is correct (see A). The second part is incorrect, but plausible, since this choice would be correct in modes 1-3, or in this mode with a lower refueling cavity water level (per TSs 3.5.2 & 3.9.5), in which two RHR Pumps are required. C -Incorrect. The first part is incorrect. Plausible, since in modes 1-4 it would be correct per TS 3.8.1. The second choice is correct (see A). 0-Incorrect. The first choice is incorrect (see C). The second choice is incorrect (see B). Page: 168 of 2n 12/14/2009 Previous NRC exam history if any: G2.2.36 2.2.36 Abil ity to analyze the effect of maintenance SJch as degraded power on the status of limiting conditions for operations (CFR: 41.10/ 43.2 / 45.13) RO 3.1 SRO 4.2 Match justification: The maintenance activity is a DG Tagged out, and it causes a degraded power source condition due to less redundancy. A scenario is provided which changes the number of available trains of decay heat removal (RHR) pumps, and the effect on limiting conditions for operations must be determined. Objective: 1 RECALL AND APPLY the LCO and APPLICABILITY for Tochnical Specifications (TS) or TRM rements, a1d the REQU I RED A CTI ONS for 1 H R or less TS or TRM rements, a1d the rei evant porti ons of BASES that D EFI N E the OPERABI LITY and APPLICABILITY of the LCO associated with the Residual Heat Removal System components and attenda1t pment aI i gnment, to i nd ude the foil owi ng (OPS-521 01 K01):
3.4.3, RCS Pressure and Temperature (PIT) Li mi ts

3.4.6, RCS Loops-MODE 4

3.4.7, RCS Loops -MODE 5, Loops Riled

3.4.8, RCS Loops-MODE 5, Loops Not Riled

3.4.12, Low Temperature Overpressure Protection (L TOp) System

3.4.14, RCS Pressure Isolation Valve (PI V) Leakage

3.5.2, ECCS -Operati ng

3.5.3, ECCS -Shutdown

3.9.4, Residual Heat Removal (RHR) a1d Coolant Circulation
-High Water Level
3.9.5, Residual Heat Removal (RHR) and Coolant Circulation
-Low Water Level
13.5.1, Emergency Core Cool i ng System (ECCS) Page: 169 of 277 12/14/2009 Question # 63 KIA G2.2.36 REFERENCE Docs AC Sources -Shutdown 3.8.2 3.8 ELECTRICAL POWER SYSTEMS 3.8.2 AC Sources -Shutdown LCO 3.8.2 The following AC electrical power sources shall be OPERABLE:
a. One qualified circuit between the offsite transmission network and the onsite Class 1 E AC electrical power distribution subsystem(s) required by LCO 3.8.10, "Distribution Systems -Shutdown";
and b. One diesel generator (DG) capable of supplying one train of the onsite Class 1 E AC electrical power distribution subsystem(s) required by LCO 3.8.10. APPLICABILITY: MODES 5 and 6, During movement of irradiated fuel assemblies. ACTIONS CONDITION A. One required offsite circuit inoperable. Farley Units 1 and 2 REQUIRED ACTION ------------------NOTE------------------- Enter applicable Conditions and Required Actions of LCO 3.8.10, with one required train de-energized as a result of Condition A. A.1 A.2.1 Declare affected required feature(s) with no offsite power available inoperable. Suspend CORE ALTERATIONS. COMPLETION TIME Immediately Immediately (continued) 3.8.2-1 Amendment No. 146 (Unit 1) Amendment No. 137 (Unit 2) RHR and Coolant Circulation -High Water Level 3.9.4 3.9 REFUELING OPERATIONS 3.9.4 Residual Heat Removal (RHR) and Coolant Circulation -High Water Level LCO 3.9.4 One RHR loop shall be OPERABLE and in operation.

NOTE--------------------------------------------

The required RHR loop may be removed from operation for s 1 hour per 8 hour period, provided no operations are permitted that would cause reduction of the Reactor Coolant System boron concentration. APPLICABILITY: MODE 6 with the water level;:::: 23 ft above the top of reactor vessel flange. ACTIONS CONDITION A. RHR loop requirements not met. Farley Units 1 and 2 A.1 A.2 REQUIRED ACTION COMPLETION TIME Suspend operations Immediately involving a reduction in reactor coolant boron concentration. Suspend loading irradiated fuel assemblies in the core. Immediately A.3 Initiate action to satisfy Immediately RHR loop requirements. 3.9.4-1 ( continued) Amendment No. 146 (Unit 1) Amendment No. 137 (Unit 2)
64. G2.3.13 001/FNP BANK/RO/C/A 3.4/3.8/Y 2008/N/3/CVR/Y Unit 1 is at 100% power, and the following conditions occurred:
Containment mini-purge supply and exhaust fans are running. R-11, CTMT ATMDS, has come into alarm. It is reading 8000 cpm. The following radiation monitors are trending up:

R-12, CTMT GAS

R-2, CTMT 155 FT

R-7, SEAL TABLE Which one of the following are the actions that the DATC is required to take for this condition lAW annunciator response procedure FH1, RMS HI-RAD?

Check pressurizer level and VCT level stable.

Secure containment mini-purge fans. B.

Ensure ALL containment mini-purge dampers have automatically closed.

Secure containment mini-purge fans. C.

Check pressurizer level and VCT level stable.

Verify ARDA has auto started. D.

Ensure ALL containment mini-purge dampers have automatically closed.

Verify ARDA has auto started. Page: 170 of 277 12/14/2009 A -Correct. First part is correct since the operator actions for all rad monitors coming into alarms states: IF RCS leakage is possible then perform actions of FNP-1-AOP-1.0, RCS LEAKAGE, per step 3.5. Second part is correct since the actions of FH1 say to IF high activity in containment is possible, THEN consider securing containment purge / minipurge (refer to FNP-1-S0P-12.2 CONTAINMENT PURGE AND PREACCESS FILTRATION SYSTEM. It also says: Perform actions of AOP-1.0 (which secures purge), per step 4.11. B -Incorrect.
First part is not correct since the containment mini-purge dampers do not close on R-11 hi rad but do close on high radiation from R-24 which monitors ctmt atmosphere when the minipurge system is running. Second part is correct -see A above. C -Incorrect. First part is correct-see A above. Second part is not correct since ARDA does not start on an R-11 signal but does auto start on R-29, 15C, 60A,B,C,D and R-14, 21,22. D -Incorrect. First part is not correct (see B). Second part is not correct -see C above ARP-1.6 Ver. 64 FH1 and FH4 FH 1 has the operator:
2. Insure that any automatic actions, associated with the alarmed channel, have occurred.
For R-11 actions it says: IF high activity in containment is possible, THEN consider securing containment purge / minipurge. AOP-1 will also have this fan secured when it is entered by procedural guidance and due to entry conditions with all the above rad monitors in alarm. Plausible since auto actions of some rad monitor does cause these auto actions to occur, just not these. Page: 171 of 277 12/14/2009 Previous NRC exam history if any: 2008 NRC Exam, this is the only question in the exam bank that meets the k/a and all NUREG 1021 Rev. 9 Supp. 1 standards. G2.3.13 2.3.13 Knowledge of radiological Slfety procedures pertaining to licensed operata duties, SJdl as response to radiation monita containment entry fuel handling responsibilities, accessto locked high-radiation aligning etc. (CFR: 41.12 / 43.4 / 45.9 / 45.10) RO 3.4 SRO 3.8 Match justification: This question asks for actions to be done by the control room operators and these actions are guided by procedure. Automatic actions of all the rad monitors are common misconceptions, and actions to take are found in the ARP. This ARP has guidance that is both generic in nature and specific to this one radiation monitor. The ARP directs actions and sends to the ReS leak AOP which directs more actions. Objective:
2. RELATE AND I DENTI FY the operational charcderistics including de3ign features, ti es ald protecti ve i nterl ocks for the components associ ated wi th the Ra::Ji ati on Monitoring System to i nclude those items in Table 4-Remoteald Local Indicationsand Controls (OPS-40305A02).

5. DEFINE AND EVALUATE the operational implications of normal / abnormal plant or equipment conditions associated with the safe operation of the Ra::Jiation Monitoring System componentsald equipment, to include thefollowi ng (OPS-40305A07):
Normal control methods

Abnormal ald Emergency Control Methods

Automatic cduation

Protective isolations

Protective interlocks

Acti ons needed to miti gate the consequence of the abnormal i ty Page: 172 of 2n 12/14/2009 Question # 64 KIA G2.3.13 REFERENCE Docs 07/02/0906:30:42 FNP-1-ARP-1.6 LOCA TI ON FH 1 OPERATOR ACTIONS 1. Check indications on radiation monitoring system console and determine which radiation monitor channel indicates high activity.
2. Insure that any automatic actions, associated with the alarmed channel, have occurred.

3. Perform the following general actions as appropriate.
3.1. Determine the source or cause of the high activity and correct or isolate as required. 3.2 Determine the validity of the high activity indication as follows: 3.2.1 Verify that the instrument is aligned for normal operation and is functioning properly. 3.2.2 IF a known problem exists such that the detector is saturated, THEN momentarily pull the affected detector's fuses (located on the front of the drawer) to clear the condition. 3.2.3 If requested to disable a remote audible alarm, refer to FNP-I-SOP-45.0, P&L 3.6. 3.2.4 Sample or survey the affected system or area as required. {CMT 0008755}. 3.3 Do not allow personnel to enter the affected area without the approval of the Health Physics Department. 3.4 IF high activity indication is due to instrument failure, THEN refer to Technical Specifications, section 3.3.3, 3.4.15 and TRM TR 13.3.4. 3.5 IF high activity indication ofRCS leakage is present AND accompanied by either decreasing pressurizer level, OR decreasing VCT level, THEN go to FNP-1-AOP-1.0, RCS LEAKAGE. 3.6 IF high activity indication of Steam Generator Tube Leakage is present, THEN go to FNP-1-AOP-2.0, STEAM GENERATOR TUBE LEAKAGE. 3.7. IF ARDA activated and not required, THEN have counting room stop the automated dose assessment per FNP-O-EIP-9.1, AUTOMATED DOSE ASSESSMENT METHOD. 3.8 WHEN radiation levels have decreased below alarm setpoint, THEN reset the appropriate HI radiation alarm on the RAD monitor drawer. Page 3 of 12 Version 61.0 07/02/09 06:30:42 FNP-I-ARP-l.6 LOCATION FHl AUTOMA TIC ACTIONS (cont) 2. ARDA will automatically start for the following conditions: 2.1 ARDA will automatically start when any of the following monitors go into alarm for two consecutive system polls one minute apart on either unit and use the latest 15 minute average monitor value to perform the calculations: Plant Vent Stack Monitors R29 (SPINO) Noble Gas 4.44e-4 Dc/ml Iodine 1.20e-6 D c/ml Particulate 4.00e-5 Dc/ml Steam Jet air Ejector RISC 27 mr/hr TDAFW Exhaust R60D 38 mr/hr Steam Generator A R60A 38 mr/hr Steam Generator B R60B Steam Generator C R60C 38 mr/hr 38 mr/hr 2.2 ARDA will also automatically start when any of the following monitors go into alarm for two consecutive system polls one minute apart on either unit. The ARDA system will use the plant Vent stack SPING latest 15 minute average monitor value to perform the calculations when these monitors activate the system: Plant Vent Stack Monitors Gas monitor R 14 Gas monitor R 21 Particulate monitor R 22 R-11 not listed .. not in circuit to start ARDA. Page 2 of 12 13,000 1800 156 Version 61.0 07/02/0906:30:42 FNP-1-ARP-1.6 RE R-IA R-1B R-2 R-3 R-4 R-5* R-6 R-7 R-8 R-9 R-I0 R-ll
LOCA TI ON FH 1 OPERATOR ACTIONS (cont) 4. In addition to the general actions perform supplementary steps indicated in the "ACTIONS" column of the following table. (Some of the radiation monitors included in the table do not input into this alarm but are included for reference).
RADIATION MONITOR REFERENCE TABLE LOCATION TYPE DETECTOR FUNCTION ACTIONS Control Room (Unit I Area G-M (W) Perform Step Panel) 4.1 Technical Support Center Area G-M (W) No inputto (Unit II Panel) R-IB this alarm Containment (155' elev) Area G-M (W) Perform Steps 4.2 Radiochemistry Lab (AB Area G-M (W) Perform Step 139') 4.3 #3 Charging Pump (AB Area G-M (W) Perform Step 100') 4.4 Spent Fuel Pool Room (AB Area G-M (W) Perform Steps ISS') 4.5 Sampling Room (AB 139') Area G-M (W) Perform Step 4.6 In-core NIS Area (CTMT Area G-M (W) Perform Steps 129', near Seal Table) 4.7 Drumming Station (AB Area G-M (W) Perform Step ISS') 4.8 SG Sample Panel (Unit II Area G-M (W) No input to Panel) (AB 139') this alarm Penetration Room Filtration APD Scint. Perform Step Discharge (AB ISS') (Victoreen) 4.10 Containment Atmosphere APD Scinto Perform Step (AB 121 ') (Victoreen) 4.11 *Technical Specification related Page 4 of 12 Version 61 .0 07/02/0906:30:42 FNP-I-ARP-l.6 A&B part #1 A&C part #2 LOCATION FHI OPERA TOR ACTION (cont) 4.8 IF R-8 in alarm THEN perform the following: 4.8.1 Announce the affected area on the public address system. 4.8.2 Have all personnel evacuate the affected area. 4.9 Step not used 4.10 IF R-10 alarms and high activity in the penetrations rooms is possible, THEN consider placing penetration room filtration in service using I-SOP-60 PENETRATION ROOM FILTRATION SYSTEM. 4.11 IF R-ll alarms, THEN perform the following: 4.11.1 IF personnel are in containment and unaware of the high activity, THEN announce the affected area on the public address system. 4.11.2 IF high activity in containment is possible, THEN consider securing containment purge 1 minipurge (refer to FNP-I-SOP-12.2 CONTAINMENT PURGE AND PREACCESS FILTRA TION SYSTEM. 4.11.3 IF RCS leakage is possible then perform actions of FNP-I-AOP-l.0, RCS LEAKAGE 4.12 IF R-12 alarms, THEN perform the following: 4.12.1 IF personnel are in containment and unaware of the high activity, THEN announce the affected area on the public address system. 4.12.2 IF high activity in containment is possible, THEN consider securing containment purge 1 minipurge (refer to FNP-I-S0P-12.2 CONTAINMENT PURGE AND PREACCESS FILTRATION SYSTEM. 4.12.3 IF RCS leakage is possible then perform actions of FNP-I-AOP-1.0, RCS LEAKAGE 4.13 IF R-13 alarms, THEN refer to FNP-I-S0P-51, WASTE GAS SYSTEM for potential problems with the waste gas system. 4.14 IF R-14 alarms, THEN perform the following: 4.14.1 IF dry storage operations are in progress, THEN perform the following, as appropriate: 4.14.1.1 IF dry storage personnel have notified OPS that the R-14 alarm is possible due to dry storage operations, THEN regard as an expected alarm. 4.14.1.2 Contact dry storage personnel AND determine if dry storage operations probably caused the R-14 alarm. LOCATION FHI Page 10 of 12 Version 61.0 07/02/09 06:30:42 SETPOINT: Variable, as per FNP-I-RCP-252 ORIGIN: Radiation Monitor Cabinet Channels R-24A or R-24B Containment Purge PROBABLE CAUSE I. High Radiation Level in the Containment Purge Exhaust Line. H4 FNP-I-ARP-l.6 LOCATION FH4 CP RE24 A OR B HIRAD 2. The radiation monitors fail to a "High Radiation" condition on loss of instrument and/or control power that will result in actuation of associated automatic functions. B&D Part #2 AUTOMA TIC ACTION 1. Isolates Containment by closing Purge Supply and Exhaust Valves l-CP-HV -3196, l-CP-HV-3197, l-CP-HV-3198A, B, C, & 0, l-CP-HV-2867C & 0 and l-CP-HV-2866C & D. OPERATOR I. Determine which radiation monitor indicates high activity.
2. Verify that any required automatic actions have occurred and if required, secure any running containment purge or mini-purge fans. 3. Notify HP personnel of alarm. 4. Implement FNP-0-EIP-9, EMERGENCY CLASSIFICATION AND ACTIONS. 5. Determine the validity of the high activity indication as follows: 5.1 Verify that the instrument is aligned for normal operation and is functioning properly.
5.2 Sample or survey the affected system or area as required.
6. Determine the source or cause of the high activity and correct or isolate as required.

7. DO NOT allow personnel to enter the affected area without the approval of the Health Physics Department.

8. IF high activity indication is due to instrument failure, THEN refer to Technical Specifications, section 3.3.6. 9. IF high activity indication ofRCS leakage is present AND accompanied by either decreasing pressurizer level OR decreasing VCT level, THEN go to FNP-I-AOP-I.O, RCS LEAKAGE.
References:
A-I77100, Sh. 309; U-258400; 0-181658; 0-181671; D-I77199; 0-177204; FSAR, Section 11.4; 0-175010, Sh. 2. Page I of I Version 61.0 04/03/09 13:20:36 FNP-I-AOP-I.O A. Purpose RCSLEAKAGE Version 18.0 This procedure provides actions for response to a minor RCS leak which is within the capacity of the normal Charging and Makeup System to maintain pressurizer level, in order to permit a controlled Reactor shutdown and RCS cool down to cold shutdown. For RCS breaks 3/8 of an inch or less, A single charging pump with letdown isolated can maintain the pressurizer level at the reactor coolant system pressure, and an ECCS actuation is not required.(FSAR 6.3.3.3 and 9.3.4.1.1.8) A 3/8 inch liquid space break is approximately a 130 gpm leak at normal operating pressure (2235 psig). This procedure is applicable in Modes 1,2 and 3. B. Symptoms or Entry Conditions I. This procedure is entered when excessive RCS leakage is indicated by any of the following:
a. Unexplained rise in charging flow b. Unexplained reduction in VCT level c. Results ofFNP-I-STP-9.0, RCS LEAKAGE TEST d. As directed from the following annunciator response procedures.
[] FNP-I-ARP-I.5 Annunciator EA2, "CHG HDR FLO HI LO" [] FNP-I-ARP-l.8 Annunciator HB2, "PRZR L VL DEV LO" [] FNP-I-ARP-I.8 Annunciator HD 1, "PRZR PRESS REL VL V 445A OR B/U HTRS ON" [] FNP-I-ARP-1.8 Annunciator HCl "PRZR PRESS HI/LO" Page 1 of25 04/03/09 13 :20:36 FNP-I-AOP-1.0 RCS LEAKAGE Version 18.0 Step Action/Expected Response Response Not Obtained -I I NOTE:
Charging flow is limited to approximately 130 gpm when FK 122 is operated in automatic.
1 [ ] [ ] [ ]
The intent of step 1 is to ensure that SI is actuated if PRZR level cannot be maintained stable above the low level heater interlock setpoint.
If possible, pressurizer level should be restored to the normal program value. A stable pressurizer level permits a controlled orderly shutdown and cooldown to cold shutdown. Maintain pressurizer level stable at or 1 Perform the following. near programmed level. Control charging flow 1.1 Verify reactor tripped. OR 1.2 IF reactor tripped, Reduce letdown flow THEN actuate SI. OR 1.3 Go to FNP-I-EEP-O, REACTOR TRIP OR IF required to maintain PRZR level, SAFETY INJECTION. THEN isolate letdown. Page Completed Page 2 of25 04/03/09 13:20:36 FNP-I-AOP-l.O RCS LEAKAGE Version 18.0 Step Action/Expected Response Response Not Obtained I I ****************************************************************************************** CAUTION: IF VCT level indication is lost due to a low level, OR IF it is suspected that the lower level tap has been uncovered, THEN the VCT level transmitters will need to be vented for reliable level indication. {2006203596}

2 Maintain VeT level greater than 20%. 2.1 Verify reactor makeup system -IN AUTOMATIC. 2.2 Control reactor makeup system in manual using FNP-I-SOP-2.3, CHEMICAL AND VOLUME CONTROL SYSTEM REACTOR MAKEUP CONTROL SYSTEM. Page Completed Page 3 of25 2 Perform the following. a) Verify charging pump suction aligned to RWST. RWST TOCHGPUMP [] QIE21LCVI15B open [] QIE21LCVI15D open VCT OUTLET ISO [] QIE21LCVI15C closed [] QIE21LCVI15E closed b) Begin unit shutdown using FNP-I-UOP-3.1, POWER OPERATION and FNP-I-UOP-2.1, SHUTDOWN OF UNIT FROM MINIMUM LOAD TO HOT STANDBY. I 04/03/09 13:20:36 FNP-1-AOP-1.0 RCSLEAKAGE Step Action/Expected Response I I I I 3 Determine RCS leak rate. -3.1 Determine RCS leak rate from CVCS flow balance. (charging flow) + (seal injection flow) -(letdown flow) -(#1 sealleakoffflow) = (RCS leak rate) 3.2 IF plant conditions are stable, THEN determine RCS leak rate using FNP-1-STP-9.0, RCS LEAKAGE TEST. 4 WHEN RCS leak rate determined, -THEN evaluate required actions using Technical Specifications. 5 WHEN RCS leak rate determined, -THEN evaluate event ciassifiation and notification requirements using FNP-O-EIP-8, NON-EMERGENCY NOTIFICATIONS and FNP-0-EIP-9, EMERGENCY CLASSIFICATION AND ACTIONS. 6 WHEN RCS leak rate greater than 50 -gpm, THEN align lA and IB post LOCA containment hydrogen analyzers for service using ATTACHMENT
1. Page Completed Page 4 of25 Version 18.0 Response Not Obtained I 04/03/09 13:20:36 FNP-I-AOP-l.O RCS LEAKAGE Version 18.0 Step Action/Expected Response Response Not Obtained -I I ****************************************************************************************
CAUTION: Since abnormal conditions may exist, nonessential personnel should not be permitted in containment until RCS leakage has been located. **************************************************************************************** NOTE:
The intent of step 7 is to provide a systematic leakage search plan. Steps 7.2 through 7.12 may be done in any order.

IF at anytime the location of an RCS leak is discovered or reported, THEN actions to isolate the leak should be taken immediately.

WHEN all leakage sources have been identified, THEN continue with step 8 and further leakage identification actions may be terminated.
7 Identify ReS leakage source. 7.1 + --= 7.2 [ ] [ ] [ ] [ ] Frequently monitor CVCS flow balance as the actions of steps 7.2 through 7.12 are taken. (charging flow) (seal injection flow) (letdown flow) (i 1 sealleakoffflow) (RCS leak rate) Check containment radiation -NORMAL. R-2 CTMT 155 ft R-7 SEAL TABLE R-l1 CTMT PARTICULATE R-12 CTMT GAS 7.2 Perform the following. 7.2.1 Consult Shift Manager to evaluate requirement for containment entry. 7.2.2 Evaluate placing CTMT sump pump handswitches in PULL-TO-LOCK to prevent overfilling the WHT. o Step 7 continued on next page Page Completed Page 5 of25 04/03/09 13:20:36 FNP-I-AOP-I.O RCS LEAKAGE Step Action/Expected Response I I 7.3 Check auxiliary building radiation-NORMAL. [] R-4 IC CHG PUMP RM [] R-6 SAMPLE RM AREA [] R-IO PRF 7.4 Check no SG tube leakage. Version 18.0 Response Not Obtained 7.2.3 Verify containment ventilation isolation. 7.2.3.l Stop MINI PURGE SUPP/EXH FAN. 7.2.3.2 Verify containment mini purge dampers -CLOSED. CTMT PURGE DMPRS MINI-2866C & 2867C FULL-3198A & 3198D [] 2866C [] 2867C CTMT PURGE DMPRS MINI-2866D & 2867D FULL-3196 & 3197 BOTH-3198B & 3198C [] 2866D [] 2867D 7.3 Perform the following. 7.3.1 Announce the hazard area using the Gaitronics System. 7.3.2 Evacuate the hazard area of essential personnel. 7.3.3 Dispatch Health Physics and Operations personnel to visually inspect accessible portions of charging, letdown, BTRS and seal injection systems using A TT ACHMENT 3, LEAK INSPECTION -121' Elevation and Above and A TT ACHMENT 4, LEAK INSPECTION -100' Elevation and Below. 7.4 Go to FNP-I-AOP-2.0, STEAM GENERATOR TUBE LEAKAGE. o Step 7 continued on next page Page Completed Page 6 of25
65. G2.3.15001/NEW/RO/MEM 2.9/3.1/N/N/3/GTO/CVR/Y A Plant Operator is assigned to use a portable RAM 100 frisker during an emergency entry. Which one of the following describes the: 1) radiation that the frisker detects and 2) the required checks prior to use lAW RCP-208, Operation and Calibration of MGP Instruments RAM 100 Count Rate Meter? A'! 1) Beta-gamma ONLY. 2) Ensure the daily response check is current and conduct a battery check. B. 1) BetaONLY.

2) Ensure the instrument responds properly to a known reference source and calibrate the instrument.
C. 1) Beta-gamma ONLY. 2) Ensure the instrument responds properly to a known reference source and calibrate the instrument. 0.1) BetaONLY.
2) Ensure the daily response check is current and conduct a battery check. Page: 173 of 2n 12/14/2009 A -Correct. Per RCP-208, Step 5.0, 5.4, & 5.5. (See below) B -Incorrect.
Beta is incorrect, but plausible, since this is memory level and confusion may exist as to which type of radiation is detected. The second part is incorrect due to the calibration not being required prior to every use. Plausible, since a calibration check is required prior to every use, but not a calibration. "Ensure the instrument responds properly to a known reference source" is correct. C -Incorrect. The first part is correct (see A). The second part is incorrect (see B). o -Incorrect. The first part is incorrect (see B). The second part is correct (see A). FNP-0-RCP-208, OPERATION AND CALIBRATION OF MGP INSTRUMENTS RAM 100 RATE METER, version 5.0. 5.0 Operation and Response Check The instrument must be response checked daily or prior to use whichever is less frequent. 5 Ensure the instrument calibration is current as indicated by the calibration sticker. 5.4 The probe will detect a beta-gamma field in CPM. 5.5 Ensure that the instrument responds properly to a known reference source. Previous NRC exam history if any: G2.3.15 2.3.15 Knowledge of radiation monitoring SJch asfixed radiation monitors and portable SJrvey per!D1nel monitoring equipment, etc. (CFR: 41.12/43.4 /45.9) RO 2.9 SRO 3.1 Match justification: This question requires knowledge of a portable survey instrument that a licensed operator may use for personnel monitoring in an emergency. Objective: OPS-30401A G2.3.5 and 2.3.15 objectives Page: 174 of 277 12/14/2009 Question # 65 KIA G2.3.15 REFERENCE Docs 10/05/09 15:43:04 FNP-0-RCP-208 4.4.3 Overflow alarm: If the displayed count rate is over 999E (999,000 cpm), the OFLO LCD's blinks on the display. IOFLOI 4.4.4 Threshold alarm: If the reading exceeds threshold value, the ALr. LCD's and the reading are displayed alternately, accompanied by an audible beep. IRLr. I Pressing the push-button mutes the audible alarm, but the ALr. LCD's and the reading are continuing to be displayed alternately, until the reading decreases to 0.75 from threshold value. In case the reading exceeds threshold value and then decreases below 0.75 of threshold value, the ALr. LCD's and the beep sound are automatically cancelled, even though the push-button has not been pressed. 4.5 Instrument is operational from _10°C to +50°C. (15°F to +122°F) 5.0 Operation and Response Check The instrument must be response checked daily or prior to use whichever is less frequent. A&D part #2 5.1 Ensure the instrument calibration is current as indicated by the calibration sticker. 5.2 Turn the instrument on by depressing the ON/OFF button. 5.3 Battery voltage is checked internally if battery voltage is low, the LCD will blink and the display will show "bAt". "bAt" alarm must be cleared out prior to use. 5.4 The probe will detect a beta-gamma field in CPM. 5.5 Ensure that the instrument responds properly to a known reference source. B & C part 2 Correct part 1 Version 4.0 10/05/09 15:43:04 FNP-0-RCP-20S 5.6 If the detector is set to alarm ensure the alarm sounds at the value indicated on the Set Point Sticker on the meter. 5.7 Clicks should be heard from the speaker during upscale readings. NOTE: Meter fluctuation is normal and is caused by the random nature of radioactive decay. NOTE: When using a low range GM tube in very high exposure fields (e.g. >2 Rlhr), the instrument may saturate. The display will show OFLO. NOTE: 5.S Push the "RESET/Mode" button and release. The reading should drop rapidly then climb back to the source reading. 5.9 If the instrument responds properly, HP Form 224 will normally be completed lAW FNP-0-RCP-201. 5.10 If the instrument does not respond properly, then remove the instrument from service lAW FNP-0-RCP-201. 5.11 The meter reading must be multiplied by appropriate correction factors to obtain the proper field intensity. The batteries should be removed from the instrument if it is to be totally inactive for a long period of time (e.g. in storage >6 months). 5.12 After use, turn off the instrument prior to storage. 6.0 Calibration 6.1 Off-site calibration The instrument may be sent off-site to a vendor for calibration in accordance with FNP-0-RCP-201. 6.2 On-site calibration Calibration of the RAM-100 is in two basic units. The base unit which includes the rate meter and display, and the detector unit which includes the amplifier and high voltage assembly CAUTION: Damage to the instrument and/or injury to personnel can occur by touching I the electrical components. 6.2.1 Base Unit 6.2.1.1 Disconnect the detector from the base unit. 6.2.1.2 Connect the base unit to a pulser via the pulser adaptor. Version 4.0 10/05/09 15:43:04 FNP-0-RCP-208 6.2.1.3 Turn on the meter. 6.2.1.4 Turn on the pulser. 6.2.1.5 Set the input amplitude to 2.7 +/- 10% volts. 6.2.1.6 Set the input Frequency to 400 cpm press the reset button on the base unit and obtain the meter reading. 6.2.1.7 Repeat for Frequency setting of 4,000 cpm and 40,000 cpm. 6.2.1.8 If the meter readings are within +/- 10% of the input frequency record the data on HP Form 240 and proceed to step 6.2.2. 6.2.1.9 If instrument does not calibrate to within +/- 10% of the reference points then remove the instrument from service in accordance with FNP-0-RCP-201. 6.2.2 Detector Calibration 6.2.2.1 Open the detector housing by removing the 3 phillips screws. 6.2.2.2 Connect the detector to the meter. 6.2.2.3 Connect a high voltage detector to the D1 cathode (See Figure 1 ). 6.2.2.4 Turn on the meter. 6.2.2.5 Measure the high voltage to the detector as the D1 cathode record "As Found" data on HP Form 240. 6.2.2.6 Expose detector to source 0515.00.00, record "As Found" data on HP Form 240. 6.2.2.7 If the meter reading is within +/- 10% of the reference source record the "As Left" data on HP Form 240 and proceed to step 6.2.2.10. If the meter reading is NOT within +/- 10% of the reference source, proceed to step 6.2.2.8. 6.2.2.8 Enter calibration mode by holding the RESET/MODE push button for 2 seconds. Then scroll using the RESET/Mode push button until the blinking display shows "CAL". 6.2.2.9 Adjust the meter reading to match the calibration source by turning R12 Calibration Factor Adjustment (see Figure 1). 6.2.2.10 If the high voltage is 875 volts +/- 5v then record this voltage in the "As Left" section on HP Form 240 and proceed to 6.2.2.13. Version 4.0 10105/09 15:43:04 FNP-0-RCP-208 6.2.2.11 If the high voltage is out of tolerance then adjust high voltage to 875 volts +/- 5v by turning R14 (See Figure 1) and record "As Left" data on HP Form 240. 6.2.2.12 Expose detector to source 0515.00.00 and record "As Left" data on HP Form 240. 6.2.2.13 Turn the meter off. 6.2.2.14 Place the detector back into the housing and tighten the 3 phillips screws. 6.2.2.15 If instrument does not calibrate to within +/- 10% of the calibration reference source then remove the instrument from service lAW FNP-0-RCP-201. 6.2.3 Affix a correction factor sticker with a correction factor of 10 to instruments using pancake type GM probes. 6.2.4 Affix a "Daily Response Check" sticker, HP Form 224 to the instrument. 6.2.5 Affix calibration sticker to instrument. Version 4.0
66. G2.3.4001/NEW/RO/MEM 3.2/3.7/N/N/3/CVR/Y Which one of the following correctly states the dose limits that Radiation Worker qualified personnel may be required to receive in an emergency lAW EIP-14.0, Personnel Movement, Relocation, Re-entry and Site Evacuation, 1) to protect valuable property and 2) to save a life? (1 ) (2) A. 5 Rem 25 Rem 10 Rem 25 Rem c. 5 Rem 100 Rem D. 10 Rem 100 Rem L-__________________________________________________________________ Page: 175 of 2n 12/14/2009 A -Incorrect.
The first part is incorrect, but plausible, since it is the 10 CFR 20 annual limit for TEDE, which is more than twice the 2 REM FNP Admin limit for normal annual dose. However, in an emergency, a radiation worker qualified individual may be required to receive up to 10 Rem to protect valuable property. The second part is correct. B -Correct. Both parts are correct. C -Incorrect. The first part is incorrect (see A). The second part is incorrect, but plausible, since it is the limit that a radiation worker can "voluntarily" receive, but cannot be required to receive over 25 Rem. D -Incorrect. The first part is correct (see B). The second part is incorrect (see C). EIP-14 v23 7.12 Emergency si tuati ons may tranocend the normal requi rement of mai ntai ni ng personnel exposure:; below 1OCFR20 limits, as noted in step 7.10. Emergency exposure:; shall be mi ni mi zed to every degree procti cabl e. Farl ey N ucl ear PI ant personnel who hewe com pi eted the onsite n:diation protection training may be required to receive an exposure up to 25 rem TEDE for the octivity and conditions deocribed below. For those same personnel to receive in excess of 25 rem, they must vol unta-i Iy agree to receive an emergency dose in excess of 25 rem, but less than 100 rem. Emergency doses received do not hewe to take into ClXX)unt the annual dose to date. Persons vol unteeri ng to receive in excess of 25 rem must be mooe fully CNVare of the risks involved. Emergency exposure limits a-e as fol lows: TEDE 10 REM 25 REM >25,<100 REM ACTIVITY PROTECTING VALUABLE PROPERTY LI FE SA. VI NG OR PROTECTION OF LARGE POPULATIONS LIFE SA.VING OR PROTECTION OF LARGE POPULATIONS CONDITION DOSE LOWER DOSE NOT PRACTICAL LOWER DOSE NOT PRACTICAL VOLUNTEERS ONLY THAT ARE FULLY AWARE OFTHE RISKS INVOLVED Limit the dose to the lens of the eye:; to 3 ti me:; the listed value. Limit the dose to other organs, including skin and extremitie:; to 10 time:; the listed value:;. Page: 176 of 2n 12/14/2009 Previous NRC exam history if any: G2.3.4 2.3.4 Knowledge of radiation expOSJre limits under normal or emergency conditions (CFR: 41.12/ 43.4 / 45.10) RO 3.2 SRO 3.7 Match justification: The question asks what are the radiation limits for the emergency conditions of protecting valuable equipment and saving a life. Objective:
6. LI ST AND I DENTI FY the individuals who ccn cuthorize re-entry into an evccuated area (0PS40501B06).
Page: 177 fA 277 12/14/2009 Question # 66 KIA G2.3.4 REFERENCE Docs 04/03/07 14:10:17 FNP-O-EIP-14.0 7.10 Farley Nuclear Plant personnel who have completed the onsite radiation protection training may be required to receive an exposure up to the following 10CFR20 limits: IOCFR20 Administrative limit limit Whole body (TEDE) -5 rem -2 rem Lens of the eyes -15 rem -6 rem Skin of the whole body -50 rem -20 rem Extremities -50 rem -20 rem Internal organs -50 rem -20 rem 7.11 Dosimetry records for potential re-entry team members are available in the Dosimetry Lab. CAUTION: EMERGENCY EXPOSURE LIMITS SHALL ONLY BE AUTHORIZED BY THE E.D. TEDE DOSE 7.12 Emergency situations may transcend the normal requirement of maintaining personnel exposures below I OCFR20 limits, as noted in step 7.10. Emergency exposures shall be minimized to every degree practicable. Farley Nuclear Plant personnel who have completed the onsite radiation protection training may be required to receive an exposure up to 25 rem TEDE for the activity and conditions described below. For those same personnel to receive in excess of25 rem, they must voluntarily agree to receive an emergency dose in excess of 25 rem, but less than 100 rem. Emergency doses received do not have to take into account the annual dose to date. Persons volunteering to receive in excess of 25 rem must be made fully aware of the risks involved. Emergency exposure limits are as follows: ACTIVITY CONDITION 10 REM PROTECTING VALUABLE LOWER DOSE NOT PRACTICAL PROPERTY 25 REM LIFE SAVING OR PROTECTION LOWER DOSE NOT PRACTICAL OF LARGE POPULATIONS >25, <100 LIFE SAVING OR PROTECTION VOLUNTEERS ONL Y THA TARE REM OF LARGE POPULATIONS FULL Y A WARE OF THE RISKS INVOLVED Limit the dose to the lens of the eyes to 3 times the listed value. Limit the dose to other organs, including skin and extremities to 10 times the listed values. Version 23
67. G2.4.13 001 /M OD/RO/M EM 4.4/4.6/N/N/3/CV R/Y Which one of the following states the MINIMUM authority by position to approve a 6 REM exposure during a Site Area Emergency?
A. The HP Supervisor ONLY. The Emergency Director ONLY. C. Either the HP Supervisor OR Emergency Director. D. Both the HP Supervisor AND the Emergency Director. A -Incorrect. ED only must approve exceeding 10CFR20 radiation exposure limits listed in step 7.10 of EIP-14. Plausible, since the HP Supervisor may authorize exceeding FNP Admin radiation limits in an emergency. B -Correct. This is required by EIP-14, Version 23.0, step 7.8. C -Incorrect. ED only must approve exceeding 10CFR20 radiation exposure limits listed in step 7.10 of EI P-14. Plausible, since the H P Supervisor, OR Emergency Director in the HP Supervisor's absence, may authorize exceeding FNP Admin radiation limits in an emergency. D -Incorrect. Plausible, since the HP Supervisor, OR Emergency Director in the HP Supervisor's absence, may authorize exceeding FNP Admin radiation limits in an emergency. It would be reasonable to assume that the HP Supervisor would be involved in and possibly required to approve the decision to exceed any radiation dose limit, and with the ED approve exceeding the 10CFR20 limit. However, the ED is soley responsible for authorizing exceeding this limit. EIP-14.0 Version 23.0 7.8 Re-Entry isthe responsibility of the Emergency Director, and ED approval to execute a re-entry. Re-entries may be authorized and executed bytheOSC Manager or Maintenance Supervi&>r, with ED approval. Approval to exceed 10CFR20 rooiation exposure limits listed in stf4) 7.10 must be approved by the Emergency Di rector. Approval to exceed plant ooministrativedose limits listed in stf4) 7.10 must be approved by the HP Supervi&>r, or the Emergency Director in the HP Supervi&>r' s absence. Page: 178 c:l2n 12114/2009 Previous NRC exam history if any: G2.4.13 2.4.13 Knowledge of crfNI roles and EOP Us:lge. (CFR: 41.10/45.12) RO 4.0 SRO 4.6 Match justification: Objective:
6. LIST AND I DENTI FY the responsibilities of individual using aprocOOure (OPS-40504A07).
Page: 179 of 277 12/14/2009 Question # 67 KIA G2.4.13 REFERENCE Docs 04/03/07 14: 10: 17 FNP-O-EIP-14.0 7.5 The re-entry guideline/log (Figures 3/4) will serve as a tracking mechanism for re-entries. One copy of the guideline will remain with the OSC and, if desired, another copy will be given to the re-entry team leader. The guideline may be photocopied, or a two-part form may be used. The re-entry guideline will be sequentially numbered. 7.6 Individuals listed on the re-entry guideline as responsible for completion of guideline items are not required to personally initial the guideline, but are responsible for ensuring that each requirement is performed and initialed by the person performing or ensuring performance of the task. 7.7 Radiological monitoring will be established for each re-entry. The following parameters will be considered when determining the degree of radiological monitoring:
Releases in progress

Dose rates, airborne and contamination levels

Stability of plant radiological conditions 7.8 Re-Entry is the responsibility of the Emergency Director, and requires verbal ED approval to execute a re-entry.
Re-entries may be authorized and executed by the OSC Manager or Maintenance Supervisor, with ED approval. Approval to exceed IOCFR20 radiation exposure limits listed in step 7.10 must be approved by the Emergency Director. Approval to exceed plant administrative dose limits listed in step 7.10 must be approved by the HP Supervisor, or the Emergency Director in the HP Supervisor's absence. 7.9 An Emergency Repair Party which functions as a re-entry team shall consist of at least two (2) persons. Version 23 04/03/07 14: 10: 17 FNP-O-EIP-14.0 7.10 Farley Nuclear Plant personnel who have completed the onsite radiation protection training may be required to receive an exposure up to the following 10CFR20 limits: 10CFR20 Administrative limit limit Whole body (TEDE) -5 rem -2 rem Lens of the eyes -15 rem -6 rem Skin of the whole body -50 rem -20 rem Extremities -50 rem -20 rem Internal organs -50 rem -20 rem 7.11 Dosimetry records for potential re-entry team members are available in the Dosimetry Lab. CAUTION: EMERGENCY EXPOSURE LIMITS SHALL ONLY BE AUTHORIZED BY THE E.D. TEDE DOSE 7.12 Emergency situations may transcend the normal requirement of maintaining personnel exposures below 10CFR20 limits, as noted in step 7.10. Emergency exposures shall be minimized to every degree practicable. Farley Nuclear Plant personnel who have completed the onsite radiation protection training may be required to receive an exposure up to 25 rem TEDE for the activity and conditions described below. For those same personnel to receive in excess of 25 rem, they must voluntarily agree to receive an emergency dose in excess of 25 rem, but less than 100 rem. Emergency doses received do not have to take into account the annual dose to date. Persons volunteering to receive in excess of 25 rem must be made fully aware of the risks involved. Emergency exposure limits are as follows: ACTIVITY CONDITION 10 REM PROTECTING VALUABLE LOWER DOSE NOT PRACTICAL PROPERTY 25 REM LIFE SAVING OR PROTECTION LOWER DOSE NOT PRACTICAL OF LARGE POPULATIONS >25,<100 LIFE SA VING OR PROTECTION VOLUNTEERS ONLY THAT ARE REM OF LARGE POPULATIONS FULLY A WARE OF THE RISKS INVOLVED Limit the dose to the lens of the eyes to 3 times the listed value. Limit the dose to other organs, including skin and extremities to 10 times the listed values. Version 23 Page: 1 Oll/HLT/LOCT/ROIC/A PROC/G2.4. 1 3111LOCTI Which of the following can authorize exposures exceeding 1 OCFR20 limits, during a Site Area Emergency? A. HP Manager. BANK Emergency Director. C. TSC Manager AND HP Manager. D. HP Manager AND Emergency Director. EIP-14.0 Version 23.0 7.8 Re-Entry is the responsibility of the Emergency Director, and requires verbal ED approval to execute a re-entry. Re-entries may be authorized and executed by the OSC Manager or Maintenance Manager, with ED approval. Approval to exceed 10CFR20 radiation exposure limits listed in step 7.10 must be approved by the Emergency Director. Approval to exceed plant administrative dose limits listed in step 7.10 must be approved by the HP Manager, or the Emergency Director in the HP Manager's absence. A. Incorrect -per the above B. Correct -per the above C. Incorrect -per the above D. Incorrect -per the above 1211112009
68. G2.4.16 001/MOD/RO/C/A 3.5/4.4/N/N/3/CVR/Y Unit 1 was at 100% power when a Large Break LOCA and a subsequent LOSP occurred.
The following conditions exist:
The crew is performing the actions of ECP-O.O, Loss of ALL AC Power.

Attempts to restore power to any 4160V bus from any source per the step, "Restoration of power to any emergency bus" have all been unsuccessful.

Core Exit Thermocouples (CETCs) read 1200°F and increasing.
Which one of the following is the required procedural flowpath? A. Continue in ECP-O.O until power is restored to at least one emergency bus. Continue in ECP-O.O until directed to transition to SAMGs (Severe Accident Management Guidelines). C. Immediately transition to SAMGs (Severe Accident Management Guidelines) from any step of ECP-O.O. D. Immediately transition to FRP-C.1, Response to Inadequate Core Cooling, from any step of ECP-O.O. Page: 180 d 2n 12/14/2009 A -Incorrect. Remaining in ECP-O.O until power is restored is incorrect, since transition to SAMGs (at step 23) would be required prior to energizing electrical busses, if they were not energized by then. Plausible, since the FRPs are not entered with loss of all AC per SOP-O.S, Emergency Response Procedure User's Guide, and are monitored for information only until power is restored. However, the SAMGs are entered to protect the containment barrier after assuming that the clad barrier is lost (CETCs > 1200°F) even with loss of all AC. B -Correct. Step 23 of ECP-O.O will direct entry into the SAMG network, but entry into SAMGs is not made until ECP-O.O has first accomplished steps which minimize DC, isolate RCS seals, dump accumulators to help cool the core, and several other things. C -Incorrect. Step 23 of ECP-O.O will direct entry into the SAMG network, and entry is not made until directed. Plausible, since in the ERG network, in most cases, and entry condition for FRPs would direct immediate entry due to the hierarchy of procedure usage. Examinee may correctly realize that the FRPs are not entered with loss of power, and the SAMGs are entered with loss of power, but incorrectly apply the immediate entry criteria for the SAMGs. D -Incorrect. The FRPs are not entered at all with a total loss of vital AC power, but this is plausible, since in the other ERPs (without loss of power), this choice would be correct. The entry condition for this procedure is met with CETCs >1200°F. SOP-0.8, Emergency Response Procedure User's Guide, Version 18.0 4.2 A ppl i cabi I i ty [of the CSFST s: FRPs] The user shoul d begi n moni tori ng the CSFST s when di rEdoo by EEP-O or upon transi ti on from EEP-O. The CSFST s are not monitoroo i niti all y bocause the ERPs are aI rea:ly di rEdi ng the initial oction roo to protEd the ba-riers. If the user enters ECP-O.O, the CSFSTs should be monitoroo for information only. The Function Restoration ProcOOuresffiSUmethat at I Em!: one train of safeguards bus.%S is availci:lle. If all AC power has been lost, ECP-O.O will provide the appropri ate octi ons to protEd the barri ers. Page: 181 of 2n 12/14/2009 Previous NRC exam history if any: G2.4.16 2.4.16 Knowledge of EOP implementation hierarchy and coordination with other SJpport procedures or guidelinesSJch operating abnormal operating and severe accident management guidelines. (CFR: 41.10/43.5/ 45.13) RO 3.5 SRO 4.4 Match justification: This question requires knowledge of the EOP hierarchy to answer correctly. The EOP heirarchy involves the FRPs being implemented as the highest priority procedures except in certain cases (early in E-O and during loss of all AC they are not implemented), also, in the event of 8AMG entry requirements, there are only a few entry points from the ERG network, and the transition is directed at the specific procedure steps. When 8AMG entry is required, it takes priority over the EOPs and FRPs. Comment: RO Knowledge of basic high level EOP priorities is being tested with this question. Objective:
1. EVALUATE plant conditions and DETERMINE if entry into (1) ECP-O.O, Loss of All AC Power; and/or (2) ECP-0.1, Loss of All AC Power Recovery, Without 81 Requiroo; and/or (3) ECP-0.2, Loss of All AC Power Recovery, With 81 Requi roo is requi roo. (0 P& 52532A 02) 2. L I Sf AN D DESCRI BE the sa:juence of or Edi ons, when and how conti nuous cdionswill beimplementoo, associatoo with (1) ECP-O.O, Loss of All AC Power; (2) ECP-0.1, Loss of All AC Power Recovery, Without 81 Requiroo; (3) ECP-0.2, Loss of All AC Power Recovery, With SI Requiroo.
(0P&52532A04)
3. ANALYZE plant conditions and DETERM INE the successful completion of any step in (1) ECP-O.O, Loss of All AC Power; (2) ECP-0.1, Loss of All AC Power Recovery, Without 81 Requiroo; (3) ECP-0.2, Loss of All AC Power Recovery, With 81 Requiroo.
(0PS-52532A07) Page: 182 d 277 12/14/2009 Question # 68 KiA G2.4.16 REFERENCE Docs 10/16/09 13 :00:27 FNP-0-SOP-0.8 4.0 Critical Safety Function Status Trees (CSFSTs) 4.1 General The ERP network is designed to protect the health and safety of the public by maintaining the fission product barriers intact. In the initial stage of ERP performance, if AC power is available, the user ensures that the automatic plant systems are functioning properly to protect these barriers. Afterwards, the CSFSTs are monitored to detect challenges to the barriers due to worsening plant conditions or equipment failure and to direct the user to an appropriate procedure. 4.2 Applicability The user should begin monitoring the CSFSTs when directed by EEP-O or upon transition from EEP-O. The CSFSTs are not monitored initially because the ERPs are already directing the initial action required to protect the barriers. If the user enters ECP-O.O, the CSFSTs should be monitored for information only. The Function Restoration Procedures assume that at least one train of safeguards busses is available. If all AC power has been lost, ECP-O.O will provide the appropriate actions to protect the barriers. 4.3 Proper Use The CSFSTs follow a logic tree format. Each CSFST has a single entry point at the left side of the page. When manually monitoring the CSFSTs, the user must enter at this point and then proceed to the right until reaching an endpoint. The endpoint will either indicate that the particular CSF is satisfied or direct the user to an appropriate procedure. The Safety Parameter Display System (SPDS) provides real time monitoring of the CSFSTs. The user should perform CSF-O to determine if SPDS is functioning properly. If so, it should be used. If not, manual monitoring is required. Version 18.0
FNP-1-ECP-0.0 LOSS OF ALL AC POWER Revision 22 Step n _21 Action/Expected Response Check PHASE B CTMT ISO not required. Response NOT Obtained 21.1 Check containment pressure -21.1 Verify PHASE B CTMT ISO. 22 _23 [] [] HAS REMAINED LESS THAN 27 psig. MLB-3 1-1 not lit MLB-3 6-1 not lit [CAl Locally monitor spent fuel pool level. (155 ft, AUX BLDG spent fuel room) 22.1 Check spent fuel pool level -GREATER THAN 153 ft. Check core exit T/Cs -LESS THAN 1200°F. 23 _Page Completed Page 35 of 40 21.1.1 Verify PHASE B CTMT ISO actuated. 21.1.2 Verify PHASE B CTMT ISO alignment. CCW FROM RCP THRM BARR [] Q1P17HV3045 closed [] Q1P17HV3184 closed CCW FROM RCP OIL CLRS [] Q1P17MOV3182 closed [] Q1P17MOV3046 closed CCW TO RCP CLRS [ ] QIP17MOV3052 closed IA TO CTMT [] Q1P19HV3611 closed (BOP) 21.1.3 Reset containment spray signal. CS RESET [] A TRN [] B TRN 22.1 Consult TSC staff to determine spent fuel pool makeup requirements. IF fifth hottest core exit Tic greater than 1200°F AND rising. THEN go to FNP-1-SACRG-1. SEVERE ACCIDENT CONTROL ROOM GUIDELINE INITIAL RESPONSE. ....--------r------------'---------------r------_---. FNP-1-ECP-O.O LOSS OF ALL AC POWER Revision 22 Step Action/Expected Response Response NOT Obtained II ************************************************************************************** CAUTION: Bus failure could result from starting loads in excess of the capacity of the power source. ************************************************************************************** _24 Check at least one train of 4160 V ESF busses -ENERGIZED. [] A Train (F & K) power available lights lit [] B Train (G & L) power available lights lit _25 Verify SW system operating. 25.1 Verify at least one SW train -HAS TWO SW PUMPs RUNNING. [] A Train (lA,lB or 1C) [] B Train (lD, IE or 1C) 25.2 Verify SW flow through at least one train of containment coolers -GREATER THAN 0 gpm. SW THROUGH CTMT CLRS INLET [] FI 3013A [] FI 3013B 25.3 Verify SW to DG BLDG valves -OPEN. SW TO/FROM DG BLDG -A HDR [] Q1P16V519/537 SW TO/FROM DG BLDG -B HDR [] QIP16V518/536 Page Completed 24 Page 36 of 40 Return to Step 11. OBSERVE CAUTION PRIOR TO STEP 11. STAY in ECP-O.O until >1200 or power is restored.
1. EC-O.OI.1.2-52532A08 015/HLTILOCT/ROICIA (LEVEL 2/3) PROC/G2.4.1611ILOCTI bank Unit 1 was operating at 100% power when a total loss of offsite anc electrical power occurred.
Given the following events and conditions: Page: 1 The crew is performing the actions of ECP-O.O, LOSS OF ALL AC POWER. Power has not been restored. The operator reports core exit thermocouples read 1200°F and increasing. Which one of the following statements correctly describes the actions the crew should take? A. Immediately go to FRP-C.1, RESPONSE TO INADEQUATE CORE COOLING. B. Remain in ECP-O.O until after power is restored to at least one emergency bus then transition to FRP-C.1. C. Complete ECP-O.O and when directed to implement monitoring CSF status trees in the appropriate recovery procedure, verify a valid RED path exists and transition to FRP-C.1. 0':'" Remain in ECP-O.O until guidance to transition to SAMGs (Severe Accident Management Guidelines) is met. A. Incorrect, FRPs are not applicable when ECP-O.O is in effect B. Incorrect. Transition to SACRG-1 is required prior to checking at least on train of 4160V ESF busses energized. C. Incorrect. SACRG-1 will be implemented at step 23 of ECP-O.O, after transition to SACRG no other procedures will implemented. D. Correct. ECP-O.O Step 23 with 5 CETs > 1200 of then go to SACRG-1. DO NOT USE WITH 52532A08 -5 12/1112009
69. WE01 EG2.2.2 001/M OD/RO/M EM 4.6/4.1/N/N/2/HBF/Y An inadvertent Safety injection has occurred on Unit 1. ESP-1.1, SI Termination, was in progress when the following conditions occurred:
MLB-1 1-1 and 11-1 lights are NOT LIT.

Pressurizer level is dropping rapidly.

SG narrow range water levels are: -1A SG 42% t. -1B SG 30% i. -1C SG 31% i.

All SG pressures are decreasing rapidly.

All Main Steam Isolation Valves (MSIVs) are open.

Ctmt pressure is 14 psig and increasing.
Which one of the following states: 1) the allowable actions to be taken per SOP-O.S, Emergency Response Procedure User's Guide, and 2) the procedure to implement IF the crew is not sure of the procedural transition? A'! 1) Close the MSIVs; 2) Enter ESP-O.O, Rediagnosis. B. 1) Isolate all AFW to 1 A SG; 2) Enter ESP-O.O, Rediagnosis. C. 1) Close the MSIVs; 2) Re-enter EEP-O, Rx Trip and Safety Injection. D. 1) Isolate all AFW to 1A SG; 2) Re-enter EEP-O, Rx Trip and Safety Injection. A -Correct. 1) Operating MSIVs is appropriate since pressure is approaching 16.2 psig automatic actuation setpoint: SOP-O.S, ver 16.0 "IF the condition is recognized in sufficient time, crews are expected to take manual actions prior to reaching the automatic setpoint for [ ... ] MSIV isolation.
2) SOP-O.S states that ESP-O.O, may be entered any time after exiting E-O, when SI is in progress OR IS REQUIRED [and no CSFs are Challenged].
Since the crew is uncertain of what action is to be taken, ESP-O.O is the appropriate action. B -Incorrect.
1) Per SOP-O.S, Step 3.3.7 Early actions may be taken since the immediate operator actions are complete.
3.3.S, Early actions to isolate Page: 183 cl2n 12/14/2009 the 1A SG may be taken since the SG is obviously ruptured, and the procedure will subsequently isolate it for the optimal recovery strategy, BUT NOT until the level is above the tubes, as indicated by the adverse numbers level of 48% minimum NR level. EEP-3, which will direct isolating all AFW at step 4, gives the minimum SG level at which the AFW flow can be isolated, and minimum level has been attained in the given conditions (1A SG level is 42%<48% minimum). Plausible, even though in this situation, the level is not high enough to secure all AFW to the SG, but at or above 31% NR level for non-adverse containment conditions or at 48% for the given adverse containment conditions (ctmt >4 psig) it would be appropriate to isolate all AFW Flow and secure feeding 1 A SG. 2) Correct See A #2 C -Incorrect.
1) See A #1 2) Re-entering EEP-O is inappropriate since E-O has already been exited; Plausible:
PER SOP-O.8 section 4.4, "IF plant conditions degrade during recovery from reactor trip without safety injection, EEP-O.O should be reentered and immediate actions performed prior to transition from ESP-O.1 to any FRP. Also, ESP-1.1 Fold out page gives direction to go to both EEP-2 (SG Fault) and EEP-3 (SGTR), and EEP-O may be chosen to provide a priority on which procedure to use for mitigation based on the EEP-O Diagnostic steps. D -Incorrect.
1) See B #1 2) See C#2 EEP-3, Steam Generator Tube Rupture, Revision 24 NOTE: [CAl Maintaining ruptured SG(s) narrow range level greater than 31%{48%} prevents SG depressurization during RCS cooldown.
4 [CAl WHEN ruptured SG(s) narrow range level greater than 31 %{48%}, THEN perform the following. FNP-O-SOP-0.8, Emergency Response Procedure User'sGuide, Version 18.0 3.7 Immediate Adions{CM T 0007770} Eerl y operator a:;ti ons shoul d not occur unti I after the i mmedi ate a:;ti ons ere verifi ed by the Shift Supervi oor. 3.8 Manual Operator Adionsand Early Operator Adions 3.8.3 Crevvs may take early operator adion when the step wi II mitigate the consequence of theeJent but not interfere with optimal reaJVery strategies. (Examples indude: socuring all but one condensate pump and ca ling for ba:;kup cool i ng to be aI i gned, taki ng manual control of Auxiliery Feedwater flow, restoring instrument air to containment, etc) The Shift Supervioor will be notified prior to the commencement of ecrly operator a:;ti on. Page: 184 ct 277 12/14/2009 Previous NRC exam history if any: 2006 NRC exam--FNP bank E-O/ESP-0.0-52530A02 012 WE01 EG2.2.2 E01 Redi agnosi s 2.2.2 Ability to manipulate the conSlie contr oIs as r equi red to oper ate the fad I ity between 91utdOlllm and desiglated povver levels (CFR: 41.6/41.7/45.2) RO 4.6 SRO 4.1 Match justification:
Requires examinee to recognize the allowable actions and expectation to be taken without procedure per the User's guide and identify proper entry into ESP-O.O, Rediagnosis.
Objective: OPS-52530A05; Analyze plant conditions and DETERMINE if actuation or reset of any ESFAS is necessary. OPS-52530A02; Evaluate plant conditions and determine if entry into [ ... ] ESP-O.O [ ... ]is required. Page: 185 of 2n 12/14/2009 Question # 69 KJA WE01 EG2.2.2 REFERENCE Docs 04/03/09 13: 18:05 FNP-0-SOP-0.8 Teamwork, like communication, is essential to effective plant operation at all times but especially during emergency operations. No individual can observe everything that is happening during a casualty. It is vital that the operating crew function as a team and maintain open communication between all team members. Each team member has a responsibility to ask questions when he does not understand something and to point out any situation in which he believes the team may be proceeding in the wrong direction. This helps to ensure that the team fully considers all aspects of the situation before reaching a decision. During an emergency event, a large number of priority tasks must be correctly performed in a limited time. In this situation, the potential for error is increased. The team can minimize this potential if each member follows procedural guidance and strictly performs each task accordingly, while communicating the status of his efforts to the other members. Team members should back each other up when possible to provide additional assurance that tasks are properly completed. 3.0 Procedure Usage 3.1 Entry Conditions There are two entry points to the ERP network. The first is if a reactor trip or safety injection occurs or is required. When this occurs the network is entered at step 1 of EEP-O, REACTOR TRIP OR SAFETY INJECTION. The second is if a complete loss of AC power to the safeguards busses occurs. For this condition the network is entered at step 1 ofECP-O.O, LOSS OF ALL AC POWER. Once the ERP network has been entered, the user is directed to other ERPs by transition steps. ESP-O.O, REDIAGNOSIS, may be entered at any time after exiting EEP-O, REACTOR TRIP OR SAFETY INJECTION, when a safety injection is in progress or is required and no red or orange path FRP is being implemented. This procedure is entered based on the user's judgment and is designed to help him determine which ERP should be implemented if any confusion develops. Each ERP has it's "Purpose" and "Symptoms or Entry Conditions" listed on the first page. This information is presented to help the user ensure that he has transitioned to the correct procedure. 3.2 Notes and Caution Statements The ERPs have been written to provide concise directed action steps for the user. For this reason, there are many cases where information in addition to action steps is provided to assist the user in proper performance of a step. If the information is needed to prevent personnel injury, mitigate the accident, prevent loss of life or prevent damage to equipment, it is placed in a caution statement. Other information is placed in a note. Version 16.0 1211 112009 15 :52 ...--------..,....-----------_.-'_.-

.....,---------...., FNP-1-ESP-0.O REDIAGNOSIS Revision 12 Step Action/Expected Response Response NOT Obtained n NOTE: This procedure should only be used if SI in progress or required.

-1 Check if any SG is not faulted. 1.1 Check pressures in all SGs -1.1 IF a controlled cooldown is in ANY STABLE OR RISING. progress. THEN proceed to Step 2. IF NOT. THEN the following applies.
IF main steam lines have NOT been isolated.
THEN go to FNP 1-EEP-2. FAULTED STEAM GENERATOR ISOLATION. OR
IF main steam lines are isolated.
THEN go to FNP-1-ECP-2.1. UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS. 2 Check if all SGs are not -faulted. 2.1 Check no SG pressure -FALLING 2.1 IF affected SG(s) NOT IN AN UNCONTROLLED MANNER OR previously isolated. LESS THAN 50 psig. THEN go to FNP-1-EEP-2, FAULTED STEAM GENERATOR ISOLATION. Page Completed Page 2 of 3 04/03/09 13:18:05 FNP-0-SOP-0.8 The immediate actions in EEP-O, FRP-S.l, and ECP-O.O will be performed, in order, by the OA TC. The UO will also perform the immediate actions in order unless directed otherwise by the Shift Supervisor. If the UO is not in the control room when an event occurs, performance of the immediate actions by the OATC alone is sufficient. When the operator(s) have finished their immediate actions and reported completion to the Shift Supervisor, the shift supervisor will verify performance of the actions using the applicable ERP. It is expected for the operator to perform manual actions to address failed ESF component actuations and to address foldout page items after the immediate actions are performed. Early operator actions should not occur until after the immediate actions are verified by the Shift Supervisor. Following verification of immediate actions, the Shift Supervisor will proceed expeditiously to implement subsequent actions. 3.8 Manual Operator Actions and Early Operator Actions 3.8.l If the condition is recognized in sufficient time, crews are expected to take manual actions prior to reaching the automatic setpoint for the following ESF actuations: Reactor Trip, Turbine Trip, SI and MSIV isolation. The determination of whether to manually initiate an anticipated automatic action would include consideration of parameter trends and applicable plant parameter values being near the setpoint. 3.8.2 Operators are expected to take manual action to address ESF components which fail to actuate when required (with the exception of starting a DG or closing the output breaker, which requires the procedure to be used to ensure load shed is verified). The Shift Supervisor should be informed as soon as possible after initiating the manual action. 3.8.3 Crews may take early operator action when the step will mitigate the consequence of the event but not interfere with optimal recovery strategies. (Examples include: securing all but one condensate pump and calling for backup cooling to be aligned, taking manual control of Auxiliary Feedwater flow, restoring instrument air to containment, etc) The Shift Supervisor will be notified prior to the commencement of early operator action. The applicable procedure step(s) will be referenced. Version 16.0 04/03/09 13: 18 :05 FNP-0-SOP-0.8 4.4 Priority The CSFSTs shall be continuously monitored in the following order of priority: S -Subcriticality C -Core Cooling H -Heat Sink P -Integrity Z -Containment ] Inventory If a red path is identified, the user will, unless specifically directed otherwise, immediately suspend the procedure in effect and transition to the specified FRP. CSFST monitoring will continue so that if a higher priority red path occurs, it will be identified. If plant conditions degrade during recovery from reactor trip without safety injection, EEP-O.O should be reentered and immediate actions performed prior to transition from ESP-O.I to any FRP. The STA should validate the need for the FRP entry while immediate actions are being performed. If an orange path is identified, the user will monitor the remaining CSFSTs to ensure that no red path exists. Unless specifically directed otherwise, he will then suspend the procedure in effect and transition to the specified FRP. CSFST monitoring will continue so that if a higher priority orange path or any red path occurs, it will be identified. If a yellow path is identified, the user is not required to transition to the specified FRP. This indicates an off normal condition that the user should be aware of, but which does not yet challenge a CSF. Implementation of a yellow path FRP is based upon operator judgement when it is determined that adequate time exists to implement it. Optimal recovery procedures (EEPs, ESPs, and ECPs) have priority over yellow path FRPs. While performing a yellow path FRP, continuous actions or foldout page items of the optimal recovery procedure in effect are still applicable and should be monitored. Concurrent procedure usage should not cause difficulties since yellow path FRPs are only performed when adequate time exists. Once an FRP has been entered due to a red or orange path, the FRP must be performed to completion unless it is preempted by a higher priority FRP. It is expected that the FRP will correct the red or orange condition before all of the operator actions are performed but the user must continue until the FRP directs a transition. In general, the performance of the critical safety functions is based on the current plant parameters. IF a red or orange path condition comes in and clears, THEN the associated FRP does not need to be performed. IF conditions degrade, THEN 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 critical safety function. Version 16.0
1. E-O/ESP-0.0-52530A02012 An inadvertent Safety injection has occurred on Unit 1. ESP-1.1, SI Termination, was in progress when the following conditions occurred:
Pressurizer level is dropping rapidly.

SG narrow range water levels are: -1A SG 85% increasing
-1 B SG 30% decreasing -1 C SG 32% decreasing
All SG pressures are decreasing rapidly.

MSIVs are open. Which one of the following would be correct if ESP-O.O, Rediagnosis, were used to determine the correct procedural transition for the above conditions?
The crew could enter ESP-O.O, Rediagnosis, to ______ _ A. verify the SG pressures are decreasing independent of a controlled cooldown and transition to ECP-2.1, Uncontrolled Depressurization of All Steam Generators. verify the SG pressures are decreasing independent of a controlled cooldown and transition to EEP-2, Faulted Steam Generator Isolation. C. close the MSIVs and verify the SG depressurization stops then transition to EEP-2, Faulted Steam Generator Isolation. D. close the MSIVs and verify the. SG depressurization stops then transition to EEP-3, Steam Generator Tube Rupture. Page: 1 of2 9/22/2009 W/E01 EK3.2 E01 Rediagnosis EK3. Knowledge of the reasons for the following responses as they apply to the (Reactor Trip or Safety Injection/Rediagnosis) (CFR: 41.5, 41.10, 45.6, 45.13) EK3.2 Normal, abnormal and emergency operating procedures associated with (Reactor Trip or Safety Injection/Rediagnosis). IMPORTANCE RO 3.0 SRO 3.9 A. This would be the correct actions if MSIVs were already closed, however the crew would not transition to ECP-2.1 since the MS lines are not isolated. B. Correct. Per step 1 of ESP-O.O with SG pressures NOT stable or rising and no controlled cooldown in progress with the MS line not having been isolated. C. Incorrect. ESP-O.O does not contain the steps to isolate the MS lines in order to create the transition path to EEP-2., and if the Depressurization was stopped, transition to EEP-2 would no longer be required. D. Incorrect. There is also a SGTR ongoing by the above conditions and the high A SG level lends credibility to this distractor. However, this transition occurs later in the steps of ESP-O.O, and would not be correct under these conditions. The priority of ESP-O.O is to send to EEP-2 first if there is a SG fault, then to EEP-3 if there is SGTR with no SG Fault, then to EEP-1 if there is no SG Fault or SGTR. FNP-1-ESP-0.0, Rediagnosis HL T-32 audit exam 2006 NRC exam KIA: Rediagnosis -Knowledge of the reasons for the following responses as they apply to the (Reactor Trip or Safety Injection/Rediagnosis): Normal, abnormal and emergency operating procedures associated with (Reactor Trip or Safety Injection/Rediagnosis). Page: 20f2 9122/2009
70. WE04EA2.2 001/MOD/RO/M EM 3.6/4.2/N/N/3/CVR/Y The Unit 1 crew has transitioned to ECP-1.2, LOCA Outside Containment.
Which one of the following correctly states: 1) a system which is isolated, and 2) the parameter used to determine if the break is isolated lAW ECP-1.2? A'! 1) RHR Cold Leg injection path. 2) RCS Pressure rising. B. 1) RHR Cold Leg injection path. 2) RCS Subcooling rising. C. 1) HHSI Cold Leg injection path. 2) RCS Pressure rising. D. 1) HHSI Cold Leg injection path. 2) RCS Subcooling rising. A -Correct. Per ECP-1.2 Steps 3.1 & 3.2. B -Incorrect. The first part is correct (see A). The second part is incorrect per ECP-1.2. Plausible, since with temperature constant, subcooling would be going up with RCS pressure going up, but the temperature and trend is not given, nor is subcooling used per ECP-1.2. C -Incorrect. First part is incorrect. Plausible, since it is a penetration into containment which is unisolated during a safety injection the same as the RHR injection to the cold leg. However, the procedure does not direct isolating this flowpath. The second part is correct (see A). D -Incorrect. Both parts are incorrect (see C & B). ECP-1.2 Version 7 Page: 186 c:J 2n 12/14/2009 Previous NRC exam history if any: WE04EA2.2 E04 LOCA Outside Containment EA2. Ability to determine and interpret the following as they apply to the (LOCA Outside Containment) (CFR: 43.5/45.13) EA2.2 Adherence to appropri ate procedures and operati on wi thi n the Ii mitati ons in the fcciI ity*s I icense and cmendments. RO 3.6 SRO 4.2 Match justification: Knowledge of the LOCA outside containment procedure is required as related to isolating the potential leak sources and the indications which are used to determine the leak is isolated. RCS leakage in a TS limitations in the facility's license, and the procedure directed leak isolation will maintain the RCS leakrate within the limits. Objective:
3. LI Sf AN D DESCRI BE the sequence of maj or ccti ons, when and how conti nuous cctionswill beimplernented, asoociatedwith ECP-1.2, LOCA Outside Containment. (OPS-52532E04)

4. EVALUATE plant conditions and DETERMINE if any 5)lstern components need to be operated whi I e performi ng ECP-1.2, LOCA Outsi de Contai nment. (OPS-52532E06)

5. ANALYZE plant conditions and DETERM INE the successful completion of any step in ECP-1.2, LOCA Outside Containment. (OPS-52532E07)
Page: 187 d 277 12/14/2009 Question # 70 KIA WE04EA2.2 REFERENCE Docs
FNP-1-ECP-1.2 LOCA OUTSIDE CONTAINMENT Revision 7 Step 11 3 Action/Expected Response Identify source of leak. 3.1 Isolate A train RHR cold leg injection path. 1A RHR HX TO RCS COLD LEGS ISO [] Q1E11MOV8888A closed RHR TO RCS HOT LEGS XCON [] Q1E11MOV8887A closed 3.2 Check RCS pressure -RISING. IC (lA) LOOP RCS WR PRESS [] PI 402A [] PI 403A 3.3 Go to FNP-1-EEP-1. LOSS OF REACTOR OR SECONDARY COOLANT. 3.4 Restore A train RHR cold leg injection path. IA RHR HX TO RCS COLD LEGS ISO [] QIEI1MOV8888A open RHR TO RCS HOT LEGS XC ON [] Q1E11MOV8887A open 3.5 Isolate B train RHR cold leg injection path. 1B RHR HX TO RCS COLD LEGS ISO [] Q1E11MOV8888B closed RHR TO RCS HOT LEGS XCON [] Q1E11MOV8887B closed Response NOT Obtained A & B part 1 C & A part 2 to step 3.4. Step 3 continued on next page. _Page Completed Page 4 of 8
1. ECP-1.2-52532E07 003/HLTIIM (LEVEL 1) PROC/W/E04EA2.2112889 Page: 1 WE04EA2.2 The crew has transitioned to ECP-1.2, LOCA Outside Containment.
Which ONE of the following parameters is used to determine if the break is isolated, in accordance with ECP-1.2? A. Pressurizer level increasing. RCS pressure increasing. C. Core exit thermocouple temperature decreasing. D. RCS subcooling increasing. DISTRACTOR ANALYSIS: A -Incorrect; PZR level alone is not sufficient indication of break isolation. B -Correct; Step 2 of ECP-1.2 C -Incorrect; Core exit thermocouple temperature decreasing is indication of heat removal not a sole indicator of break isolation. o -Incorrect; RCS subcooling increasing is indication of heat removal and possible pressure increase but not a sole indicator of break isolation. 2006 NRC exam Turkey Point 2002 NRC Exam 12/14/2009
71. WE05EK1.1 001/NEW/RO/MEM 3.8/4.1/N/N/4/HBF/Y Given the following plant conditions for Unit 1 :
A Train is On Service.

The Operators have implemented FRP-H.1, Response to Loss of Secondary Heat Sink.

RCS feed and bleed criteria was met and a manual Safety Injection was initiated lAW FRP-H.1.

1 C Charging pump is tripped.

PRZR PORV, PCV-445A, will not open. Which one of the following describes the MINIMUM action(s) required to provide adequate core cooling? A. Open one PORV. B. Open all Reactor Vessel Head vents. Open one PORV AND Open all Reactor Vessel Head vents. D. Open one PORV AND Open all Reactor Vessel Head vents, AND place 1 B Charging pump on B train and start 1 B Charging pump. A -Incorrect.
Both PORVs are required per FRB-H.1 of the background documents for FRP-H.1, Loss of heat sink Function Restoration Procedure. B -Incorrect. One PORV and all Head vents must be open to provide an adequate heat sink if one PORV cannot be opened. C -Correct. As stated in FRB-H.1 ver 2.0 , for ERP step 17 basis (below), the function provided by the second PORV capacity is cooling: "[ ... ] If both PRZR PORVs are not maintained open, the RCS may not depressurize sufficiently to permit adequate feed of subcooled SI flow to remove core decay heat. If core decay heat exceeds RCS bleed and feed heat removal capability, the RCS will repressurize rapidly, further reducing the feed of subcooled SI flow and resulting in a rapid decrease of RCS inventory. [ ... ] IF a low pressure water source can not be aligned [to at least one intact SG], a SG should not be depressurized in order to minimize the risk of tube creep rupture [ ... ]." Safety Capacity: From TS B2.4.1 0, each safety is capable of 345,000 Ib/hr Assuming that each HHSI pump can deliver 600 gpm each at 2485 psig, then the following calculation (not adjusting for Temperature correction which lowers the gal/Ibm #) below still provides sufficient relief capacity to prevent integrity failure of the RCS due to overpressure conditions. = 538560lbs / hr min gal hr Page: 188 d 277 12/14/2009 FNP-1-FRP-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK, Revision 26 17.3 Open both PRZR PORVs. 17.3 Perform the following. 17.3.1 Open all available PORV's. 17.3.2 Open reactor vessel head vent valves. RX VESSEL HEAD VENT OUTER ISO o 01B13SV2213A o 01B13SV2213B RX VESSEL HEAD VENT INNER ISO 001 B13SV2214A o 01 B13SV2214B D -Incorrect. Starting a second pump is not required, since one pump will deliver the required flow. Plausible, since two PORVs are required, and it may be incorrectly assumed that in this case both HHSI pumps are required. Starting a second pump impacts pump discharge pressure, and may be construed to improve the overall pump head capacity. Although starting a second centrifugal pump aligned in a parrallel configuration with another causes discharge pressure to rise, the shutoff head of both pumps remains unchanged. This action only alters the current operating point on the pump curve, and does not improve the overall capability of the pumps, as RCS pressure approaches the Safety valve setpoint, SI flow is continualy degraded. Previous NRC exam history if any: WE05EK1.1 E05 Loss of Seconda-y Heat Sink EK1. Knowledge of the operational implications of the following concepts as they apply to the (Lossai' Secondary Heat Sink) (CFR: 41.8/41.10,45.3) EK1.1 Components, and function of emergency RO 3.8 SRO 4.1 Match justification: Knowledge of the PORV'S function and capacity requirements of the PORV during bleed and feed operations Objective:
2. RELATE AND I DENT I FY the operational charcderistics induding design features, and protective interlocks for the components as.c:ociatEd with the Pressurizer System, to i nd ude the components found on Fi gure 3, Pressuri zer and Pressuri zer Rei i ef Tank (OPS-40301 E02). Page: 189 c:l277 12/14/2009 Question # 71 KIA WE05EK1.1 REFERENCE Docs
09/28/07 13:12:01 RESPONSE TO LOSS OF SECONDARY HEAT SINK Plant Specific Background Information FNP-O-FRB-H.l Unit 1 ERP Step: 17 Section: Procedure Unit 2 ERP Step: 17 ERG Step No: 15 ERP StepText: Establish RCS bleed path. ERG Step Text: 1. Establish RCS Bleed Path; 2. Verify Adequate RCS Bleed Path Purpose: I. To open all PRZR PORVs to establish an ReS bleed path. 2. To verify that an adequate RCS bleed path is established and, if not, to establish alternative bleed path or cooling methods. Basis: 1. The operator ensures that the pressurizer block valves are open and opens both pressurizer PORVs to establish an RCS bleed path. These valves must be maintained in the open position until secondary heat sink is restored. Once the pressurizer PORVs are open, the RCS will depressurize and the charging/SI pumps and/or high-head SI pumps will deliver subcooled flow to the RCS. This will provide adequate RCS heat removal until flow can be established to the steam generators to restore secondary heat sink. 2. After manually opening the pressurizer PORVs, the operator should check that both pressurizer PORVs are maintained in the open position. Ifboth valves are maintained open, sufficient RCS bleed flow exists to permit ReS heat removal. If both PRZR PORV s are not maintained open, the Res may not depressurize sufficiently to permit adequate feed of sub cooled SI flow to remove core decay heat. If core decay heat exceeds RCS bleed and feed heat removal capability, the ReS will repressurize rapidly, further reducing the feed of subcooJed SI flow and resulting in a rapid decrease of RCS inventory. Although only one open PRZR PORV may not be sufficient to maintain adequate RCS bleed flow, the operator should maintain one PRZR PORV open, if possible, and open all RCS high point vents to provide additional bleed path capability. In addition, the operator should align any available low pressure water source to the SO(s). The operator should then attempt to open a steam generator PORV for at least one intact SO and depressurize that SO to atmospheric pressure to inject the low pressure water source to restore secondary heat removal. If a low pressure water source can not be aligned, a SO should not be depressurized in order to minimize the risk of tube creep rupture that can occur following a severe accident where the SO tubes are subjected to high Res temperatures and large primary-to-secondary pressure differences. It should be noted that RCS inventory depletion will occur from the open single PRZR PORV, the PRZR safety valves, and high point vents as the steam generator(s) is being depressurized to atmospheric pressure. Knowledge: 0 The operator should verify that the pressurizer PORVs do not automatically close following release of the control board switches. If the pressurizer PORVs do automatically close due to a spring return to auto switch, the operator should manually maintain the control board switches in the open position. 0 The operator may observe increasing pressurizer level after the pressurizer PORVs are opened. Eventually the pressurizer may become water solid with water relief occurring through the pressurizer PORVs. 48 of 77 Version: 2.0 Pressurizer Safety Valves B 3.4.10 B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.10 Pressurizer Safety Valves BASES BACKGROUND Farley Units 1 and 2 The pressurizer safety valves provide, in conjunction with the Reactor Protection System, overpressure protection for the RCS. The pressurizer safety valves are totally enclosed pop type, spring loaded, self actuated valves with backpressure compensation. The safety valves are designed to prevent the system pressure from exceeding the system Safety Limit (SL), 2735 psig, which is 110% of the design pressure. Because the safety valves are totally enclosed and self actuating, they are considered independent components. The relief capacity for each valve, 345,000 Ib/hr, is based on postulated overpressure transient conditions resulting from a complete loss of steam flow to the turbine. This event results in the maximum surge rate into the pressurizer, which specifies the minimum relief capacity for the safety valves. The discharge flow from the pressurizer safety valves is directed to the pressurizer relief tank. This discharge flow is indicated by an increase in temperature downstream of the pressurizer safety valves or increase in the pressurizer relief tank temperature or level. Overpressure protection is required in MODES 1, 2, 3, 4, and 5; however, in MODE 4, with one or more RCS cold leg temperatures 325°F, and MODE 5 and MODE 6 with the reactor vessel head on, overpressure protection is provided by operating procedures and by meeting the requirements of LCO 3.4.12, "Low Temperature Overpressure Protection (L TOP) System." The upper and lower pressure limits are based on the +/- 1 % tolerance requirement (Ref. 1) for lifting pressures above 1000 psig. The lift setting is for the ambient conditions associated with MODES 1, 2, and 3. This requires either that the valves be set hot or that a correlation between hot and cold settings be established. The pressurizer safety valves are part of the primary success path and mitigate the effects of postulated accidents. OPERABILITY of the safety valves ensures that the RCS pressure will be limited to 110% of design pressure. The consequences of exceeding the ( continued) B 3.4.10-1 Revision 0
72. WE08EK1.1 001/NEW/RO/C/A 3.5/3.8/N/N/2/HBFNER 5 EDITORIAL Given the following plant conditions for Unit 1 :
A failure has occurred on TI-412, Loop A TAVG, resulting in a constant output equivalent to 561°F.

A failure has occurred on TI-422, Loop B T AVG, resulting in a constant output equivalent to 570°F.

The Reactor is manually tripped. Which one of the following states the protective feature that will prevent a Pressurized Thermal Shock condition from developing, with no operator actions? A. Low Main Steam Pressure SI. B. P-4, Reactor Trip Interlock.
C. Main Steam Line Isolation on High Flow. D'!" Main Steam Line Isolation on Low Pressure. Sequence of events: With 2/3 TAVG failed above 543,547, and 554 with a RX Trip, then the following occurs: 1) cooldown initiates on Rx Trip controller of steam dumps (ARMED by P-4) and Median Tavg is failed at 561°F = dumps STAY open trying to try to lower median Tavg to 54rF. 2) Main Feedwater Isolation on P-4 wi 2/3 Protection Tavg channels .::;.554°F does NOT occur to secure feed from SGFPs, so more cooldown due to ever feeding occures until P-14, SG Hi Hi level is reached. 3) P-12 will not actuate on 2/3 TAVG.::;. 543°F resulting in all steam dumps staying open. 4) even though there is excessive steam flow (Dumps + SGFP + other steam loads), there will be no high steam flow MSIV isolation since there is no P-12 actuation on LO LO T AVG. 5) Stm pressure falls to 585 psig ( rate compensated) a) S I is actuated b) MSLlS is actuated A -Incorrect. Low Main Steam pressure Si will actuate, however a Pressurized Thermal Shock (PTS) condition is initiated by cooldown. An SI actuation can cause pressure to be maintained high during a cooldown event which will actually complicate, NOT PREVENT, plant conditions creating a PTS condition, particularly if RCS pressure is rapidly restored after a significant cooldown. PTS is a HIGH pressure condition with a significant COOLDOWN. Page: 190 of 2n Plausible: This failure will result in a Low MS pressure SI actuating on Low MS header pressure of 585 psig. 12/14/2009 B -Incorrect. The FW isolation due to P-4 and LO TAVG does not occur due to the failures, but if it did it would limit the severity of this event by isolating the Feed Water flow; Plausible: Main Feedwater Isolation is initiated to prevent excessive cooldown of the reactor or to lessen the severity of the transient overall. (A 181007 pg 2-26). C -Incorrect. The failures prevent the LO LO TAVG signal from occurring. Therefore only 1 (High steam flow) of the 2 parts (HIGH flow concurrent with LO-LO TAVG) of this signal will actuate. o -Correct. WOG FRG-P.1 Plausible: the purpose of this protective function is to back up the Low Steam line pressure MSLlS, for conditions when MSLlS has been blocked. This signal would, if it could actuate, limit the effects of the uncontrolled steam release from the SG and thereby, limit the cooldown. MS pressure will drop due to the TRIP controller of the Steam dumps causing an open signal to the steam dumps, and P-12 will not actuate to close the dumps at 543°F. By isolating the MSIVs, the steam release is stopped and the cooldown would be stopped at approximately 487 of Tcold which is <100 of from the 100% TCoid value of approximately 54rF. Therefore PTS condition would be averted by this actuation signal. An event or series of events which lea:Jsto a relatively rapid and severe recdor vessel downcomer cool down can resul tin a thermal shock to the vessel wall that may I eaj to a small flaN, which may alrea:Jy exist in the vessel Wall, growing into a larger crack. The growth or extension of such aflaN may lecd, in some cases [ ... ], to a loss of vessel integrity. NOTE TO EXAMINER: Previous NRC exam history if any: WE08EK1.1 E08 Pressurizoo Thermal Shock EK1. Knowledge of the operational implications of the following concepts as they apply to the (PresaJrized Thermal Shock) (CFR: 41.8/41.10,45.3) EK 1.1 Components, capa::ity, and function of emergency &ystems. RO 3.5 SRO 3.8 Match justification: The operational implication of the failure of the Protection Tavg signal on preventing PTS must be understood to select the correct answer. In the conditions given, the PTS is still prevented, but not by the same method as would occur if Tavg was operable. In this case, PTS is prevented only by the MSLlAS which closes MSIVs on Low Steam line pressure. Three other signals which would normally prevent or mitigate a PTS event are disabled by the failure (Hi Steam Flow/Lo Lo Tavg, Page: 191 of 2n 12/14/2009 low steam line pressure.
The given failure of the Median Tavg circuit (component), results in failure of several Reactor protection functions (function of emergency systems) which are either directly or indirectly involved with preventing, mitigating, or terminating a PTS challenge.
P-4 coincident with 10 tavg--overfeed= overcooling= PTS challenge MSLlS--Low pressure ---main function is for CNMT protection but also functions to isolate a break when down stream of the MSIV--eliminating or terminating the C/D. (indirectly protecting from an oversteam event = overcooling=PTS challenge). MSLlS--HIGH FlOW with LO-LO TAVG---backup to the MSLlS low pressure for low power or shutdown modes of operation when MSLlAS-LP is blocked. SI--provided as valid distractor since SI actually can complicate a PTS condition by allowing rapid repressurization of the RCS, following a large cooldown. (PTS and/or Cold REPRESSURIZATION concern) but is required for core cooling. Objective:
1. RECALL AND DESCRIBE theopa-ation and function of the foil owing ra:dor trip signals, pa-missives, control interlocks, and engineerEd safeguards cctuation signals asoociatEd with the Ra:dor ProtECtion System (RPS) and EngineerEd Safeguards Features (ESF) to indude setpoint, coincidence, rate functions (if any), reset features, and the potenti aI consequences for improper condi ti ons to i nd ude those items in the following tables (OPS-52201 107):
T abl e 1, Ra:dor T ri p Si gnal s

T abl e 2, Engi neerEd Safeguards Features Actuati on Si gnal s

Table5, Permissives

T abl e 6, Control i nterl ocks 5. DEFI NE AND EVALUATE the operational implications of abnormal plant or equipment conditions asoociatEd with the operation of the Ra:dor ProtECtion System (RPS) components and equi pment to i nd ude the foil owi ng (OPS-52201I 09). Page: 192 eX 277

Normal Control Methods

A bnormal and Emergency Control Methods

Automatic cctuati on i nd udi ng setpoi nt ( exampl e SI, Alase A, Alase B, M SLIAS, LOSP, SG level)

Acti ons neEdEd to miti gate the consequence of the abnormal i ty 12/14/2009 Question # 72 KIA WEOSEK1.1 REFERENCE Docs
2. DESCRIPTION An event or series of events which leads to a relatively rapid and severe reactor vessel downcomer cool down can result in a thermal shock to the vessel wall that may lead to a small flaw, which may already exist in the vessel wall, growing into a larger crack. The growth or extension of such a flaw may lead, in some cases (where propagation is not stopped within the wall), to a loss of vessel integrity.
The objective of Function Restoration Guideline FR-P.l is to prevent the growth of a flaw and, in the event the limits set forth are exceeded, provide specific actions which appropriately restrict operation to prevent further challenges to vessel integrity. Two separate types of events lead to entry into this guideline: a Pressurized Thermal Shock Events Several possible transients can be hypothesized which will produce rapid and extensive temperature decreases in the RCS cold leg(s) and, by inference, also the reactor vessel downcomer region. The rate and extent of cool down determine whether entry into this guideline is on a RED or ORANGE priority. The actions in this guideline attempt to stop the cooldown, i.e., stabilize temperature, and also decrease RCS pressure to reduce the pressure stress component of total stress in the reactor vessel wall, partially offsetting the large thermal stress created by the rapid cool down. o Cold Overpressure Events For this type of event, entry on an ORANGE priority is warranted if RCS pressure has exceeded the Cold Overpressure Protection Limit, and ReS cold leg temperature is sufficiently low that vessel ductility is reduced. There is little or no thermal stress associated with this event, so the benefit in using this guideline comes from the prompt RCS pressure reduction actions which supplement the Cold Overpressure Protection System. FR-P.l Background HFRPIBG.doc 2 HP-Rev. 2, 4/30/2005 FNP Units 1 & 2 REACTOR PROTECTION SYSTEM A-181007 for the required engineered safety features lines. Phase B isolation is initiated by containment pressure High-3 (27 psig) or by manual actuation ( using 2/4 Containment Phase B Isolation/Containment Spray Actuation handswitches). The Containment Ventilation Isolation isolates the containment atmosphere from the environment to limit the release of radioactive fission products in the event of an accident. This function is actuated on the completion of the SI logic, high radioactivity levels in the purge exhaust, or by manual initiation of either Phase A Containment Isolation or Phase B Isolation/Containment Spray Actuation. (References 6.1.022, 6.4.007, 6.4.015, 6.7.012, 6.4.080) 3. Main Steam Line Isolation Isolation of the Main Steam lines limits the effects of an uncontrolled release of steam either inside or outside the containment. For a break upstream of the isolation valves (MSIV) in the steamlines, valves closure will limit the release to the blowdown of the one affected steam generator. A break downstream of the valves is limited to the depressurization of the pipe volume downstream of the valves. This results in a rapid termination of the event and significantly reduces the mass lost from the secondary. The Main Steam Line Isolation is initiated by the following:
a. High steam line flow with low-low T avg , 112 steam flow channels above setpoint (40% of full steam flow between 0-20% load and increasing linearly to 110% at full load) on 2/3 steam lines with Tavg:::; P-12 b. Low steam pressure;
585 psig on 2/3 S.G. c. High-2 containment pressure; 2 16.2 psig on 2/3 d. Manual. By closing each MSIV by operating individual hand switches. (References 6.1.022, 6.4.007, 6.4.015, 6.7.012) 4. Main Feedwater Isolation and Turbine Trip The Main Feed Line Isolation is initiated to prevent excessive cooldown of the reactor or to lessen the severity of the transient overall. The following signals are utilized to initiate the Main Feed line Isolation

2-26 Rev. 10 I ROD BOTTOM SIGNAL ANY FULL LENGTH ROD FROM ROD POSITION INDICATION SYSTEM 1 RPsrSDo9.DVG T AVG. T AVG l::,. T b. T 6. T LOOP 1 LOOP 2 LOOP 1 LOOP 2 LOOP 3 I I I I I I I J I I I I I I I I I I I I I I I I I , I I I I I I I I I 1 1'7'\ I I I t-0 t-0 C-5 C-4 C-3 C-2 C-J TURBINE OVER-OVER-HIGH FLUX HIGH FLUX POVER <15% PO'WER TEMP. (/4) <l/2) <TURBINE IMPULSE /":,. T /). T <PO'w'ER INTERMEDIATE CHAMBER PRESSURE) (2/3) (2/3) RANGE> RANGE> (SHEET 15) (SHEET 5) <SHEET 5) <SHEET 4) (SHEET 4) -=IY'u: : : :!IT\;;.. I -----'Y -t-y i-----* y 1 I I I I I I I j t j j : ,lJ :: ,1I: I I I I I ,I : &3 &3 :: .&3 : : (NOTE 5) (NOTE 5> : : (NOTE I I I I _________ . ____ $ 1.1,1 NOTE";, 1. ALL CIRCUITS ON THIS SHEET ARE NOT REDUNDANT RODS IN RODS OUT I I I I I I I I I I I I FlXGI MANUAL I ROD SPEED : I I
ANALOG ROD SPEED SIGNAL AUTO-MANUAL SELECTOR S\JlTCH rULL LENGTH CONTROL BANKS TO V LEVEL CONTROL I (NOTE 3) I ---.. r--I (NOTE 3) I (SHEET 1D
--r----------t I I / I I : : (( I I: : ,,,,,V:, i N __________ + ____ .,-____ : : :: :
:: : V I I I I I .---.; .---i .---i I t---,
j"I, ([) LCD c1..CD 1.. A"i AA AA A A"A LOW' LJ-LD LO'" LO-LO LOV LO-LO LOV LO-LO BANK A BANK B BANK C BANK D C-J! 2. KOT MAY VARY INVERSELY PROPORTIONAL TO LOAD "'!TH A F"IXED LIMIT, DR MAY VARY IN T"'O DISCRETE STEPS \.lITH BREAK POINTS AT 30-50% AND &0 TO 80% TURBINE LOAD 3. THE SUMMER OUTPUTS HAVE FIXED MANUALLY ADJUSTABLE UPPER LIMITS 4. THE ROD DIRECTION BISTABt.ES ItSS';08C ARC 'ENERGIZE TO ACTUATE' 5. ALARM: AND ALARM 3 MUST HAVE REF"LA$H CAPABILITY. ROD CONTROLS AND BLOCKS FIGURE F -2 SHEET 9 FNP UNITS 1 8. 2 RPS-FSD A-181007 H '0 " d,l" "', r , l , !'t-l '-I " fJ\ , II V ' . LJ TRIP All r(£O\IAT(R PUHPS O(j'T[ 3 a. 5} "","AL RESEr <t(lT( 4) TUR!UN( TR!P (SK£T 9 r---------------------------------------------------- ([(DVATER "EOVATER FEEOVATER STE" STEAM '[[])VATER flO\{ I Fl .. OV FlOV Fi.OIt rlOV rt.O'J rlOV i I ;"'1 " i I I I I I " ,-,-+-'t-311-S-9 t ': 1 l+"'C3QS :"L: I , I j-,+-'t-311-S STEAM GENERATOR M3 STEAM GEN,.,.r"" .? I I : 1 ::: If: :: i I ::: '1-: 1 I I I I I f +L J; :: : -Xi""' :::: 1::: .., J I J I " .... I I I L.. ( ....... I, I: : 2:: CONSTANT, I' I LJ CONSTANT ",:. >. p I l)f1 tVJ ,;---r: 1VJ tV? ' -'---!It-' 'V5 M " I K311V+¥")
! KlOV+'t3lS)

U--1 K311V+'t3lS
--_ ,: r----'-----.;1...+ ,: '-r::;:l+ : ___ I I U I I ... ________ J i I e m1 '" I ___ 3 .,. -,-, \lil'-J I m ** m '[ + I ,-1..-,--, _, DISCHARGE PRESS. 1+"tJ .. S : I
-u-'--._.......l I I 1 I t I I
... ----0 .. ----0 ,,----0 rl-AU-ro-_.L- .......... --,j
I L?kHO : AUXfP. : AI)X-r.P.

UJ;<FP. I r--0 I I I I , I r I stTPOINT I STATU)t I STATICH I 'lATIDN : l", _ + , L ____ -@-_____ J . L ________ , I I I I I I I I , I 1 :----0 , rtCDVAT[I?
POMP SPEED (,t,LL I I I I I I I I I , I I I L---0 I t---0 I I . I I I -u-HOTESI I. 2, 3. I I , : I I ! I I I I , I I I I I I I I I Ti I I rCR EACH TItAIN. S. lk1 P(.RoIp DISCHAR(£ VALVES 6. [P(H/5H.)T INDICATION fOR EACH rEEn\lAT£R VALVE IN COIITRtIL II!lDt 7. Tf1£ STEAK IZl£RATLJi! NARRQ\I RAHGE lEV(l SICiHAI.. fOR nIDlJAT(R: CONTROl.. IS fH£ I'(DI"'" (MIDDlE> SIGNAL OF LEVEL CtwHl.t a. AN AlARM IS PRQVlDED TN TI-£ CI}HI'ID.. TO INDICATE THAT EITHER TI-( MSS <MEDIAN SIGNAL :EL[CTQR) Pil'J(R SUF?L Y HAS rAILED CR It rAUl r IN Tk'£ M'iS KAS OCCURRED. , I I1JtlJI..ATE rEED'JATER MAIN VALVE rCV-Hij eNUiE 1£) I MODULATE F,(EDlafATER BrPASZ VALvE O(!n: 6) I :<IDtR..IoTE f((O\IATER MAIN rCv-:3S lNOtE b: i MOOUlAT( r-(EOW'!\T(R ... SS VAlVE <NOTE 6> § RPSF2-13.n .... G [OVY2000 I CJS-02 I (M]T[ U 1 t I HOOUlATE fEED\lATER BY"ASS VAL V£ O(li£ 6> FEED\JATER CONTROL AND ISOLA TION F'IGURE F -2 SHEET 13 I REV. 9 03:25:10 PM 11116101 CLHAYES [ llH.D liNG llIP 3 OVERTEMl'ERATURE l!. T ROCl1I! OOP (St£ET 2> C-4 DVERPOIIER l!. T ROCl1I! TRJP (SKIT 2> lJ<<ERVIl.TAGE RCP BUSSES ____ ElUSI BUSZ IUS 3 AIJXIl.W!' FErDllATER PIlI' (Sf£ET 14> D (]J) m ? /"" REIU<DNIT I FLOII LDOP I I' I lITl£Rl IY I R(ACT!R TRIP ££T 2> 0IlT£ 3> FLOII LOOP 2 ROC"" TRJP <SKIT 2> B,II!1l£R' IYI@ IV J tm£RS 11UI'", OOTE 3> 0IlT£ 3> BY I DT>£RS FLOII CiioP-3-----+---REV. 2 PRDlARY COOLANT SYSTEM TRIP SIGNALS FIGURE F-2 SHEET 5 REFERENCE U-166235 & U-207294 .'t' I ..... QTI-\'t .. , r::i --___ L---_ -l C[M)(HSER SIGNAL I I STEAN\ QUNlP.r .... I I IN"iERl..oc.l( SE\.E.C.TOct C\RC.UU'l.T\ON Wli:rER I p * .,. I I &MTCH ( .... 0'1'& 3) P\.)o"jP II RE,tr.ClOf\ nit> .,.. LR'E '!.W\:.. :"CNr;ERS -r:1 .. s I ME.mAN (' '' ........ SCPontr p.MP GOI'J1'1<'O\.Wo,OOE Sn:,o.M "£ADEQ I I I I I I BY I I I ("""aT 5) I L I 1 1. I 1 S'i'£,IiM ... u;oEi:l "''"' .... ll(l*r.;."') Ltp$ -----, , I I I I I I I I I I 1-----' p' c-=9 1 1 I I I .'1'" I BV O'Il&IUj I I t MODULA.TE MOOUL .... '1Wt LOOP 1 I ATMOSPHI!CIIC. ATt.\OSPtHoR\C f I '*-LV! "'-LVE RI!!,l'Ic::F "",we.. Ii'" OTIIUS r. 5TEAM CUMPI$8LOCKEO eo..,. 8LOCI(IIolG AIR. TO nut OUMP ANovlf'n'u*.:a T\.I£ OI .... TrE REDUNDANT LOGIC OUTPUT OPERATES Z SOI.£NOIO VENT VALVES IN SER\ES TO REOUNOiJHLV INTERLOCK 1)£ Alit lUI: BETWEEN EACH VAlVE DIAPl<<AGl\ f.HO 1lS ASSOCIATED PilSITIOf£R. Tl£ SOLENVlU VALVES ARE DE-ENERGIZEO TO VENT CAUSING llE. IMIN .ou"p VJ.lVE TO CLOSE IN FIVE SECONOS. 2.. CIRCUITRY 00 litiS StlID IS NOT REDUNDANT EXCEPT \n£R[ INDICJ.lED I':EDlINOfo}/t. l S[l£CTct. swntH WrtH
~~FOLLOWING 3 STEAM DUMP IS PERMITTED.~~
AVG INTERlOCK IS BYPASSED fOR lOW T ING. SPRING RErur.N TO ON PUSITIOH. .. I 6 -----"1 + *t:-@..::;:.-;_ i ,-------- "',ORO-----I-.J "TO l'I-It. fOlLO'W'ltoiCa SEQUENCE_ Cf'F-Sl£m DUMP IS NOT p(r;MlmD AND RESET T AVG BYPASS. Tt£ R£DUHDAHT 1N1"£RLOCK stK£CTCR S\lITDI CONSISTS of"T'NO C.ONTRO .... S. ON 'TWi. CO","-ROL FOR 'i:.JIo,.CHTR,..IN.
~~1. '-ffO~~
.. 'aN ........ IHP\JT", Fl'IOM C:OME FROM ""'P'S To M'EE.'-"SING\-,,- F,..,II..\l'R,.'I! CilITli.Rl0",a. I I (Non: j) 5. THE CONO'EK'$'i .......... A.II.A't\.'io l.06lC. !II T"tPIC. ..... L. \lo\PI..IiM!.NT,,-TION tot",,\,, '&E CI'F1=1iRS,"'T. 6.,a.,u."TIiMPt!:AATURIii ON.TK'So N-lO"t'\lAe\K'E IMP\J\.'!:oE C.\o\ .... Me.ER, PRE'5.suiI,.E 61ST)..,!U
~~1i IPB-"17A * ...~~
~~1b ... c."u,..,.!.-:~~
1. LlG.HTS SHOULD BE PROVIDED IN TH E" CONTROL ROOM FOR e: .... CH DUMP VILVE TO INC4C.I'I.TE WHEN THE VI<UIE 15 FULL V CLOSED OR flJU.y OPEN. REV. 4 7/98 JOSEPH M, FAR\.EY STEAM DUMP CONTROL NUCLEAR PLANT UNI"-1 AND UNIT 2 FIGURE F-2, SHEET 10

73. WE11 EK1.3 001/NEW/RO/C/A 3.6/4.0/N/N/3/CVR/Y A Dual unit LOSP with a Unit 2 Large Break LOCA has occurred and the following conditions occurred:
CTMT Pressure is 6 psig.

EEP-1.0, Loss of Reactor or Secondary Coolant, is in progress.
~~At 1000: WA2, 1-2A DG GEN FAULT TRIP, comes into alarm. At 1020: the following alarms have just come in:~~
CF3, 2A OR 2B RHR PUMP OVERLOAD TRIP

CH2, RWST LVL A TRN LO

CH3, RWST LVL B TRN LO Which one of the following is: 1) the correct status of Unit 2 emergency recirculation capability, and 2) the action(s) that the applicable procedure(s) direct? A. 1) One train ONLY of emergency recirc capability has been lost. 2) Transfer to Cold Leg recirc AND Containment Spray recirc at this time. B. 1) One train ONLY of emergency recirc capability has been lost. 2) Transfer to Cold Leg recirc, but do NOT transfer to Containment Spray recirc at this time. C"! 1) Both trains of emergency recirc capability have been lost. 2) Verify both Containment spray pumps secured, AND minimize HHSI flow to the minimum required to remove decay heat. D. 1) Both trains of emergency recirc capability have been lost. 2) Verify both Containment spray pumps AND HHSI pumps are secured while attempting to restore at least one train of emergency recirc. A -Incorrect.
The first part is incorrect A). The second part is incorrect, but plausible. If the fist part was correct, the second part would be correct except for the "at this time". The CS recirc line up is not begun at the RWST LOW level alarm (12.5 ft) even thought the ECCS recirc alignment is. The CS recirc alignment would be commences at the RWST LOW LOW level alarm (4.5 ft), and not "at this time". B -Incorrect. Both trains of emergency recirc are lost. A Train is indicated lost due to the only available A train DG tripped alarm WA2 (A train RHR, HHSI, and CS pumps are deenergized). CF4 has been stated as "in alarm" for consistency with a loss of A train RHR flow due to A train losing power. The B train recirc capability Page: 193 d 2n 12/14/2009 has been lost due to the two alarms together: CF3 & CF5 indicating that the 8 Train RHR pump has tripped. Plausible, since improper diagnosing either train with the indications given would lead applicant to believe one train was still available. The second part is incorrect, since no trains of recirc are available, but plausible. If the first part was correct, the second part would also be correct (Le. transfer to one train of Cold leg recirc and leave the Containment spray system in the injection mode until the LO LO RWST level at 4.5 ft, then transfer the CS system to recirc). C -Correct. 80th trains of Emergency recirc capability have been lost (see A). The high level actions of this procedure that a RO is required to know is that flow from the RWST is minimized and makeup to the RWST is maximized. For this scenario, one Containment spray pump is pumping the RWST water to the containment where it is unavailable to cool the core. ECP-1.1 will direct securing the Containment Spray pump and throttle the HHSI flow to the minimum required to cool the core. Commence makeup to the RWST is also required, but was not included in the correct answer for brevity. The answer is still correct without every action that will be completed. D -Incorrect. The first part is correct (see A). The second part is incorrect, since some minimum HHSI flow will be maintained. Plausible, since all pumps would be secured at the RWST LOW LOW level alarm which comes in at 4.5 ft. Confusion may exist as to the difference between the action and setpoint for the RWST Lo alarm and the Lo Lo alarm. Also, attempting to restore at least one train of emergency recirc is directed by the procedure. OPS-52531G, ESP-1.3, TRANSFER TO COLD LEG RECIRCULATION, ESP-1.4, TRANSFER TO SIMULTANEOUS COLD AND HOT LEG RECIRCULATION, lesson plan: Major Action Categories in ESP-1.3 The maj or ccti on categori es a-e di s:;uSSEd bel ow in more detai I. 1. Align ECCSfor recirculation.
~~2. Align CTMT spray for recirculation.~~
OPS-52532D, ECP-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION, ECP-1.3, LOSS OF EMERGENCY COOLANT RECIRCULATION CAUSED BY SUMP BLOCKAGE Major Acti on Categori es in ECP-1.1 A high level summary of thecctions performed in ECP-1.1 is given in the form of major cction categori es. 1. Conti nue attempts to restore ECR. 2. IncrEBSe/conserve RWST level. 3. Initiatecool down to cold shutdown.
4. Depressurize RCSto minimize RCSsubcooling.

5. Try to aid makeup to RCS from alternate source. 6. Depressuri ze SGs to cool down and depressuri ze RCS. 7. M ai ntai n RCS hoot removal. Page: 194 d 277 12/14/2009 Previous NRC exam history if any: WE11 EKI.3 E11 Loss of Emergency Coolant Rocirculation EK 1. Knowledge of the operational implications of the following concepts as they apply to the (Loss of Emergency Coolant Recirculation) (CFR: 41.8/41.10/45.3)
EK 1.3 Annund ators and condi ti ons i ndi cati ng si gnal s, and rernOOi aI cdi ons 8S9)Ci atOO wi th the (Loss of Emergency Coolent Rocirculation). RO 3.6 SRO 4.0 Match justification: The first part of the question is written to present various Alarms and require the applicant to determine that the status of emergency recirc capability in that both trains of emergency recirc is lost. The second part of the question requires the applicant to know what the remedial actions for the loss of recirc are (high level RO required knowledge). Objective:
1. EVALUATE plant conditions and DETERM I NE if entry into (1) ECP-1.1, Loss of Emergency Cool ant Roci rcul ati on; end/or (2) ECP-1.3, Loss of Emergency Cool ant Rocirculation, CaJsed by Sump Blockage is ra:JuirOO. (OPS-52532D02)

2. EVALUATE plant conditions end DETERM I NE if any &ystern components need to be operatOO while performing (1) ECP-1.1, Loss of Emergency Coolant Rocirculation; (2) ECP-1.3, Loss of Emergency Coolent Rocirculation, CaJsed by Sump Blockage. (OPS-52532D06)
Page: 195 of 277 12/14/2009 Question # 73 KIA WE11EK1,3 REFERENCE Docs 06/26/06 18:04:34 FNP-I-ARP-l.3 LOCATION CF3 1A pump lost power 20 SETPOINT: Variable Current/Time mins before L.!:'lJ ORIGIN: 86 Relay (DF-09) Control Circuit, lA RHR Pump or 86 Relay (DG-09) Control Circuit, IB RHR Pump PROBABLE CAUSE 1. 1 A or 1 B RHR Pump overloaded
2. lA or IB RHR Pump electrical or mechanical fault AUTOMA TIC ACTION 1 A or 1 B RHR Pump Trip. OPERATOR ACTION lA OR IB RHRPUMP OVERLOAD TRIP I. Check indications and determine which RHR pump has tripped. 2. IF the NON-AFFECTED pump is NOT running, THEN start the pump in accordance with FNP-I-AOP-12.0, RESIDUAL HEAT REMOVAL SYSTEM MALFUNCTION.

3. Refer to FNP-I-AOP-12.0, RESIDUAL HEAT REMOVAL SYSTEM MALFUNCTION.

4. Notify appropriate personnel to determine and correct the cause of the RHR Pump trip. 5. Return the RHR System to normal operation as soon as possible.

6. Refer to Technical Specifications, Sections 3.4.6, 3.4.7, 3.4.8, 3.5.3, 3.9.4 and 3.9.5, for LCO Requirements.
~~References:~~
A-I77100, Sh. 168; D-175041; D-I77193; A-I 77048, Sh. 260 & 274; Technical Specifications. Page 1 of 1 Version 27.0 06/26/06 18:04:34 SETPOINT: 12'7" +/- I" above Tank Bottom ( 150,000 Gallons) ORIGIN: Level Transmitter QIFI6LT-501 through a comparator card bistable designated LSL503 in BOP Cabinet 1. PROBABLE CAUSE 1. RWST in use for Safety Injection purposes.
2. RWST in use for Refueling purposes.

3. Failed Level Transmitter.
~~AUTOMATIC ACTION NONE OPERA TOR ACTION FNP-I-ARP-I.3 LOCATION CH2 RWST LVL A TRN LO 1. IF an ECCS actuation signal is present, THEN refer to FNP-I-ESP-1.3, TRANSFER TO COLD LEG RECIRCULATION.~~
2. Determine actual tank level as indicated by LI-4075A & B, on the MCB OR the local level indicator on the side of the RWST. 3. IF an ECCS Actuation Signal is NOT present OR the tank is NOT being used for Refueling, THEN notify appropriate personnel to determine and correct the cause of the alarm. 4. IF required, THEN restore RWST level to normal per FNP-I-SOP-2.3, CHEMICAL AND VOLUME CONTROL SYSTEM REACTOR MAKEUP CONTROL SYSTEM, section 4.2.3. 5. Refer to Technical Specification 3.3.3 for LCO requirements.
~~References:~~
A-I77100, Sh. 177; A-170750, Pg. 95; D-173497; Technical Specifications Page 1 of 1 Version 27.0 06/26/06 18:04:34 SETPOINT: 12'7" +/- I" above Tank Bottom (150,000 Gallons) ORIGIN: Level Transmitter QIFI6LT502 through a comparator card bistable designated LSL504 in BOP Cabinet K. PROBABLE CAUSE 1. RWST in use for Safety Injection purposes.
2. R WST in use for Refueling purposes.

3. Failed Level Transmitter.
~~AUTOMATIC ACTION NONE OPERA TOR ACTION FNP-I-ARP-l.3 LOCATION CH3 RWSTLVL B TRN LO 1. If an ECCS actuation signal is present, THEN refer to FNP-I-ESP-I.3. TRANSFER TO COLD LEG RECIRCULATION.~~
2. Determine actual tank level as indicated by LI-4075A & B, on the MCB OR the local level indicator on the side of the RWST. 3. IF an ECCS Actuation signal is NOT present OR the tank is NOT being used for Refueling, THEN notify appropriate personnel to determine and correct the cause of the alarm. 4. IF required, THEN restore RWST level to normal per FNP-I-SOP-2.3, CHEMICAL AND VOLUME CONTROL SYSTEM REACTOR MAKEUP CONTROL SYSTEM, section 4.2.3. 5. Refer to Technical Specification 3.3.3 for LCO requirements.
~~References:~~
A-I77IOO, Sh. 178; A-I70750, Pg. 95; B-170058, Sh. 72; D-173497; Technical Specifications Page 1 of 1 Version 27.0
FNP-l-ECP-l.l LOSS OF EMERGENCY COOLANT RECIRCULATION Revision 27 A. Purpose This procedure provides actions to restore emergency coolant recirculation capability. to delay depletion of the RWST by adding makeup and reducing outflow. and to depressurize the RCS to minimize break flow. B. Symptoms or Entry Conditions I. This procedure is entered when emergency coolant recirculation capability is lost; from the following:
~~a. FNP-l-EEP-l.~~
~~LOSS OF REACTOR OR SECONDARY COOLANT. step 14. when cold leg recirculation capability cannot be verified.~~
~~b. FNP-l-ESP-l.3.~~
~~TRANSFER TO COLD LEG RECIRCULATION. step 7. when at least one flow path from the containment sump cannot be established or maintained.~~
~~c. FNP-l-ECP-l.2.~~
LOCA OUTSIDE CONTAINMENT. step 3. when a LOCA outside containment cannot be isolated. Page 1 of 51
FNP-1-ECP-1.1 LOSS OF EMERGENCY COOLANT RECIRCULATION Revision 27 Step Action/Expected Response II _10 Evaluate containment spray requirements. 10.1 Check containment spray pumps -ALIGNED TO RWST. RWST TO 1A(1B) CS PUMP [] Q1E13MOV8817A open [] Q1E13MOV8817B open 10.2 Determine number of containment spray pumps required based on the Table below. RWST LEVEL CONTAINMENT PRESSURE GREATER THAN 54 PSIG GREATER THAN BETWEEN 12.5 FT 27 PSIG AND 54 PSIG LESS THAN 27 PSIG GREATER THAN 54 PsrG BETWEEN BETWEEN 4.5 FT 27 PSIG and 54 PSIG and 12.5 FT LESS THAN 27 PSIG LESS THAN 4.5 FT --10.3 Establish required number of started containment spray pumps. _Page Completed Page 5 of 51 Response NOT Obtained 10.1 IF containment spray pumps aligned to the the containment sump, THEN proceed to Step 12. FAN COOLERS RUNNING IN EMERGENCY MODE --0, 1 2, 3 4 ----1, 2 3, 4 --SPRAY PUMPS REQUIRED 2 2 1 0 0 2 1 0 0 0
FNP-1-ECP-1.1 LOSS OF EMERGENCY COOLANT RECIRCULATION Revision 27 Step Action/Expected Response II NOTE: Step 21 is a continuing action. ___ 21 [CAl Check SI termination criteria. 21.1 Check REACTOR VESSEL LEVEL .
IF any RCP started. THEN check REACTOR VESSEL LEVEL greater than 72% UPPER PLENUM .

IF no RCP started. THEN check REACTOR VESSEL LEVEL greater than 0% UPPER PLENUM. Response NOT Obtained 21.1 Proceed to step 27. Step 21 continued on next page. ___ Page Completed Page 20 of 51
.....------....,...--------_._-,-,----:._----------,----------. FNP-I-ECP-l.l LOSS OF EMERGENCY COOLANT RECIRCULATION Revision 27 Step Action/Expected Response Response NOT Obtained 1\ NOTE: TABLE 1 provides mlnlmum SI flow to remove decay heat vs. time elapsed after shutdown. 21.2 Check SUB COOLED MARGIN MONITOR -GREATER THAN 66°F {9S0F} SUBCOOLED IN CETC MODE. 21.2 Establish minimum SI flow. 21.2.1 II charging pump suction aligned to RHR, THEN stop all CHG PUMPs. 21.2.2 Verify both RHR PUMPs stopped. 21.2.3 Open miniflow valve for available charging pump. lA(lB,lC) CHG PUMP MINIFLOW ISO [] QIE21MOV8109A [] QIE21MOV8109B [J QIE21MOV8109C 21.2.4 Open common miniflow isolation valve. CHG PUMP MINIFLOW ISO [] QIE21MOV8106 21.2.S Verify RWST to charging pump valves open. RWST TO CHG PUMP [] QIE21LCVllSB [] QIE21LCVllSD 21.2.6 Close RHR supply to A AND B train charging pump suction. lA(lB) RHR HX TO CHG PUMP SUCT [] QIEIIMOV8706A [] QIEIIMOV8706B Step 21 continued on next page. _Page Completed Page 21 of Sl
FNP-1-ECP-1.1 LOSS OF EMERGENCY COOLANT RECIRCULATION Revision 27 Step Action/Expected Response Response NOT Obtained II _Page Completed 21.2.7 Start CHG PUMP with miniflow valve open. 21.2.8 Maintain core exit Tic temperature stable or falling.
II core exit Tic temperatures rising AND started charging pump aligned to A train, THEN establish A train SI flow. HHSI TO RCS CL ISO [] Q1E21MOV8803A open [] Q1E21MOV8803B open

IF core exit Tic temperatures rising AND started charging pump aligned to B train, THEN establish B train SI flow. CHG PUMP RECIRC TO RCS COLD LEGS [] Q1E21MOV8885 open 21.2.9 Open and close HHSI isolation valves to control SI flow to keep core exit Tic temperatures stable or falling. HHSI TO RCS CL ISO [] Q1E21MOV8803A
[] Q1E21MOV8803B CHG PUMP RECIRC TO RCS COLD LEGS [] Q1E21MOV8885 Step 21 continued on next page. Page 22 of 51 r-------------,---------------------.-----------------------------r-----------------, FNP-I-ECP-l.l LOSS OF EMERGENCY COOLANT RECIRCULATION Revision 27 Step Action/Expected Response II -22 Reset safeguards Signals. 22.1 Verify PHASE A CTMT ISO RESET. [] MLB-2 1-1 not lit [] MLB-2 11-1 not lit 22.2 Verify PHASE B CTMT ISO RESET. [] MLB-3 1-1 not lit [] MLB-3 6-1 not lit 23 Establish instrument air to containment. 23.1 Check ID 4160 V bus -ENERGIZED. 23.2 IF ID 4160 V bus energized. THEN proceed to step 23.4. 23.3 Establish power to lA 600 V LC emergency section loads. 23.3.1 Verify open BKR EA08-1. 23.3.2 Verify closed BKRs ED08-1 and EA09-1. Response NOT Obtained 21.2.10 Consult TSC staff to determine if additional options are available or needed to further minimize SI flow and conserve RWST inventory. 21.2.11 Proceed to step 27. 23.1 Proceed to step 23.3. Step 23 continued on next page. _Page Completed Page 23 of 51
FNP-1-ECP-1.1 LOSS OF EMERGENCY COOLANT RECIRCULATION Revision 27 Step Action/Expected Response Response NOT Obtained II ************************************************************************************** CAUTION: To ensure adequate supply voltage to all class IE loads and to meet short circuit analysis constraints. only one air compressor. 1C (preferred) or 1A. may be powered from the diesel generator. One air compressor will consume 0.16 MW of diesel generator load. ************************************************************************************** 23.4 Verify 1C air compressor in service. 23.4.1 Verify 1C air compressor handswitch in AUTO after START/RUN. 23.4.2 Verify 1C air compressor started. 23.4 Align 1A air compressor for service. a) Verify 1C air compressor handswitch in OFF. b) Verify SI -RESET. [] MLB-1 1-1 not lit [] MLB-1 11-1 not lit ************************************************************************************** CAUTION: IF offsite power is lost after sequencer is reset. THEN manual actions may be required to restart safeguards equipment.
* * * * * * * * * * * * * * * * * * * * * * *

* * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * -A,-* * * * * * * * * * * * * * * * * * * *

___ Page Completed c) Reset B1F sequencer by depressing the ESS STOP RESET pushbutton on the sequencer panel. (139 ft. AUX BLDG A train SWGR room) d) Place BKR DF13 SYNCH SWITCH in MAN. e) Close BKR DF13 (IF 4160 V bus tie to 1H 4160 V bus) . f) IF 1H 4160 V bus energized.
THEN energize 1G 600 V LC from normal supply. [] BKR DH01 closed [] BKR EG02-1 closed g) Start 1A AIR COMPRESSOR. Step 23 continued on next page. Page 24 of 51
FNP-1-ECP-1.1 LOSS OF EMERGENCY COOLANT RECIRCULATION Revision 27 Step Action/Expected Response Response NOT Obtained II 23.5 Check INST AIR PRESS PI 4004B -GREATER THAN 85 psig. 23.5 Perform the following. NOTE: The intent of this step is to regain control of critical air operated components including PORVs and atmospherics. Based on plant conditions and availability of manpower. the applicability. priority and performance of the following actions is at the discretion of the Shift Supervisor. 23.5.1 Restore air pressure .
~~Verify proper air compressor operation using FNP-1-S0P-31.0.~~
~~COMPRESSED AIR SYSTEM.~~
~~IF 2C air compressor available.~~
THEN align 2C air compressor to Unit 1 using FNP-1-S0P-31.0. COMPRESSED AIR SYSTEM. 23.5.2 IF instrument air NOT restored. THEN align nitrogen supply to PORVs using FNP-1-S0P-62.1. BACK-UP AIR OR NITROGEN SUPPLY TO THE PRESSURIZER POWER OPERATED RELIEF VALVES. 23.5.3 IF instrument air NOT restored. THEN align emergency air supply to atmospheric relief valves and/or TDAFWP using FNP-1-S0P-62.0. EMERGENCY AIR SYSTEM. Step 23 continued on next page. ___ Page Completed Page 25 of 51
FNP-1-ECP-1.1 LOSS OF EMERGENCY COOLANT RECIRCULATION Revision 27 Step Action/Expected Response II 23.6 Check instrument air to containment. IA TO CTMT [] MLB-3 1-2 NOT lit IA TO PENE RM PRESS LO [] Annunciator KD1 clear --24 Stop SI pumps. 24.1 Verify both RHR PUMPs -STOPPED. 24.2 Verify only one CHG PUMP -STARTED. _Page Completed Page 26 of 51 Response NOT Obtained 23.5.4 IF instrument air NOT restored because 1G 600 V LC is deenergized, THEN energize 1G 600 V LC from IF 600 V LC using FNP-1-S0P-36.3, 600, 480 AND 208/120 VOLT AC ELECTRICAL DISTRIBUTION SYSTEM. 23.5.5 WHEN instrument air pressure restored, THEN perform step 23.6. 23.5.6 Proceed to step 24. 23.6 Align instrument air to containment. (BOP) IA TO PENE RM [] N1P19HV3825 open [] N1P19HV3885 open IA TO CTMT [] Q1P19HV3611 open
FNP-1-ECP-1.1 LOSS OF EMERGENCY COOLANT RECIRCULATION Revision 27 Step Action/Expected Response Response NOT Obtained II I I I 25 Isolate HIlSI flow. -25.1 Verify charging pump miniflow valves -OPEN. 1A(1B.1C) CHG PUMP MINIFLOW ISO [] Q1E21MOV8109A [] Q1E21MOV8109B [] Q1E21MOV8109C CHG PUMP MINIFLOW ISO [] Q1E21MOV8106 25.2 Close HHSI isolation valves. HHSI TO RCS CL ISO [] Q1E21MOV8803A [] Q1E21MOV8803B -26 Establish normal charging. 26.1 Manually close charging flow control valve. CHG FLOW [] FK 122 Step 26 continued on next page. _Page Completed Page 27 of 51 ---FNP-I-ECP-l.l LOSS OF EMERGENCY COOLANT RECIRCULATION Revision 27 Step Action/Expected Response Response NOT Obtained II I I I I 26.2 Verify charging flow path aligned. 26.2.1 Verify charging pump discharge flow path -ALIGNED. CHG PUMP DISCH HDR ISO [] QIE21MOV8132A open [] QlE21MOV8132B open [] QIE2lMOV8l33A open [] QlE21MOV8133B open CHG PMPS TO REGENERATIVE HX [] QIE21MOV8107 open [] QIE21MOV8108 open 26.2.2 Verify only one charging line valve -OPEN. RCS NORMAL CHG LINE [] QlE21HV8l46 RCS ALT CHG LINE [] QIE21HV8147 26.3 Establish desired charging flow using charging flow control valve. CHG FLOW [] FK 122 _Page Completed Page 28 of 51
FNP-1-ECP-1.1 LOSS OF EMERGENCY COOLANT RECIRCULATION Revision 27 Step 11 NOTE: _27 Action/Expected Response Step 27 is a continuing action. [CAl Verify RCS makeup flow -ADEQUATE. 27.1 Check REACTOR VESSEL LEVEL.
IF any RCP started. THEN check REACTOR VESSEL LEVEL greater than 72% UPPER PLENUM.

IF no RCP started. THEN check REACTOR VESSEL LEVEL greater than 0% UPPER PLENUM. Response NOT Obtained 27.1 Increase RCS makeup flow to maintain required REACTOR VESSEL LEVEL.

CHG FLOW [] FK 122 adjusted

HHSI TO RCS CL ISO [] Q1E21MOV8803A open [] Q1E21MOV8803B open

CHG PUMP RECIRC TO RCS COLD LEGS [] Q1E21MOV8885 open Step 27 continued on next page. _Page Completed Page 29 of 51
FNP-1-ECP-1.1 LOSS OF EMERGENCY COOLANT RECIRCULATION Revision 27 Step Action/Expected Response II 27.2 Check core exit TiCs stable or falling. Page Completed Page 30 of 51 Response NOT Obtained 27.2 Increase RCS makeup flow to maintain core exit TiCs stable or falling.
CHG FLOW [] FK 122 adjusted

HHSI TO RCS CL ISO [] Q1E21MOV8803A open [] Q1E21MOV8803B open

CHG PUMP RECIRC TO RCS COLD LEGS [] Q1E21MOV8885 open
~~74. WE12EK2.1 001/NEW/RO/C/A 3.4/3.7/N/N/3/HBF/Y Unit 2 is in Mode 3, preparing to open the MSIVs after warm-up of the Main Steam lines is complete using SOP-17.0, Main and Reheat Steam.~~
ReS Temp is 54rF.

All SG MSIV Bypass valves are open.

Steam header pressure and each SG pressures are approximately equal. The UO opens the MSIVs, and immediately after opening all MSIVs:

A large Steam Break occurs in the Turbine Building.

The MSIVs would not close when the MCB handswitches were placed in the CLOSE position.
~~Which one of the following describes:~~
1) the position of the MSIV Bypass valves, and 2) the system response to placing each SG MSIV-TEST switches to TEST? A. 1) The MSIV Bypass valves are open. 2) The MSIVs will be driven fully closed with air; B. 1) The MSIV Bypass valves closed. 2) The MSIVs will be driven fully closed with air; C. 1) The MSIV Bypass valves are open. 2) The MSIVs will be partially closed with air; 1) The MSIV Bypass valves are closed. 2) The MSIVs will be partially closed with air; Page: 196 of 277 12/14/2009 A -Incorrect.

1) The MSIV Bypass valves are interlocked with the MSIVs such that they close immediately when the MSIV is not Closed. Further, if MSLlAS is satisfied ( 585 psig rate sensitive) the BYPASS valves will have shut. Plausible:
~~The MSIVs are normally shut AND are not operated by the MCB MSIV "TRIP" handswitch directly. These valves are equipped with their own handswitches.~~
2) The test Cylinder is only capable of moving the MSIV disc off its backseat, it is incapable of driving the valve fully closed. System forward flow impacting the disc (swing-check) will force the valve shut. B -Incorrect.

1) this is correct. The MSIV bypass valves will be closed immediately upon the MSIV Closed limit switch not being satisfied.

2) See A #2 C -Incorrect.

1) See A #1. 2) This portion is correct. D -Correct.

1) The MSIVs are interlocked such that when the MSIV is no longer closed, the bypass valves close. 2) The TEST cylinder provides enough air force to move the MSIV disc off its backseat, in addition to the spring that should normally move the valve off the backseat, but should not cause the disc to enter the flow stream. It is the weight of the disc and the flow of steam which results in the MSIV closing. Page: 197 of 277 lnT tnl.III ** _$T(AIIII iSOlaTION VAU'E UIIIGrt , lIIC' 12/14/2009 Previous NRC exam history if any: WE12EK2.1 E12 Uncontrolled Depressurization of all Steam Generators EK2. Knowledge of the interrelations between the (Uncontrolled Depressurization of all Steam Generators) and the following: (CFR: 41.7 /45.7) EK2.1 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
RO 3.4 SRO 3.7 Match justification: the Uncontrolled depressurization --STEAM break with NO successful MSIV closure. Manual features, and interlocks--- MSIV Bypass valves are interlocked with MSIV position to ensure that the MSLlS can complete its function. These actions are those required by ECP-2.1 in an attempt to terminate the uncontrolled depressurization. Objective:
6. DEFINE AND EVALUATE the operational implications of normal / abnormal plant or equipmEnt conditions a.s&>eiated with the safe operation of the Main and Reheat Steam System componEnts and equipmEnt, to indudethefollowing (OP&40201A07):
Normal control methods

Abnormal and EmergEnCY Control Methods

Automatic cctuation induding setpoint (example SI, Phase A, Phase B, MSLlAS, LOSP, SG level)

Protective isolations such as high flow, low pressure, low level including setpoint

Protective interlocks

Actions needed to mitigate the consequEnce of the abnormality Page: 198 ct 2n 12/14/2009 Question # 74 KIA WE12EK2.1 REFERENCE Docs 07/02/09 06:40: 16 FNP-2-S0P-17.0 NOTE: When an MSIV is opened, its associated bypass valve will automatically close. The upstream and downstream MSIV for at least one loop must be opened before proceeding to the other loops to maintain the bypass flowpath.
~~4.3.10 4.3.11 4.3.12 WHEN stearn header pressure AND individual stearn generator pressures are approximately equal, THEN open the MSIVs.~~
Q2NIIHV3369A

Q2NIIHV3370A

Q2N II HV3369B

Q2N IIHV3370B

Q2N II HV3369C

Q2NIIHV3370C WHEN the MSIV s reach the fully open position, THEN verify that the associated bypass valves automatically close. WHEN desired, THEN unisolate main stearn drain pot level control valves by performing the following:

Open N2N II V905A 2A MS LINE DRN POT TO COND ISO

Open N2NIIV905B 2B MS LINE DRN POT TO COND ISO

Open N2N II V905C 2C MS LINE DRN POT TO COND ISO

Open N2N II V905D 20 MS LINE DRN POT TO COND ISO 4.3.12.1 Clear the admin tracking item initiated in step 4.3.1. Version 50.0
~~75. WE16EK2.1 001/BANK/RO/C/A 3.0/3.3/N/N/2/HBF/Y A Large Break LOCA has occurred on Unit 2, and the following conditions exist:~~
R-27A and B, CTMT HI RANGE, indicates 3 Rem/hr .

RE-11, CTMT PART, and RE-12, CTMT GAS, on the Integrated Plant Computer (I PC) shows an initial upscale followed by a slow trend towards background levels. Which ONE of the following describes the reason for the observed trend on RE-11 and RE-12 towards a background count rate? RE-11 and RE-12 are isolated from containment directly from a ___ signal. A'I Phase A B. Phase B C. Safety Injection D. Containment Ventilation Isolation A. Correct, CTMT ATMOS TO R-11 /12 01 E14MOV3660,3657
&3658 are closed by a containment isolation -'T' signal. B. Incorrect, Containment phase B occurs at a higher pressure and does not affect the R-11/R-12 valves. C. Incorrect, CTMT ATMOS TO R-11/12 01 E14MOV3660,3657 &3658 close on a 'T' signal not an'S' signal. D. Incorrect, Containment ventilation isolation signal is generated by a MANUAL Phase A or B signal, any signal that generates an SI, HI-HI rad on RE-24A/B and does not affect the isolation valves for R-11 /12. EEP-O, ATTACHMENT 2, Revision 38 Page: 199 of 277 12/14/2009 Previous NRC exam history if any: N/A WE16EK2.1 E16 High Containment Rooiation EK2. Kn(Mfledge of the interrelations between the (H ig, Containment Radiation) and the foll(Mfing: (CFR: 41.7/45.7) EK2.1 Components, ald functi ons of control ald safety systems, i ncl udi ng i nstrumentati on, signals, interlocks, failure modes, ald automatic and malual features. RO 3.0 SRO 3.3 Match justification: The interrelation between the given high radiation condition in in containment during an accident and the function of the safety system and the automatic features which isolate the high radiation from the public must be understood to answer this question. High containment radiation is indicated on R-27A & B as well as R-11 & 12 due to the LOCA, and then R-11 & 12 are isolated due to the Automatic ISOLATION SIGNALS. This meets entry criteria for the FNP-1-FRP-Z.3, Response To High Containment Radiation Level (yELLOW PATH: Both CTMT RAD NOT LESS THAN 2 R/hr.) per FNP-1-CSF-0.5 CONTAINMENT Revision 17 Objective: RMS-40305A07 Page: 200 of 2n 12/14/2009 Question # 75 KIA WE16EK2.1 REFERENCE Docs -_ .. _----12111/200915:5 r FNP-I-EEP-O REACTOR TRIP OR SAFETY INJECTION Revision 37 Step Action/Expected Response Response NOT Obtained II I I I I ATTACHMENT 3 PHASE A CONTAINMENT ISOLATION NOTE:
ATTACHMENT 3, FIGURE 1 provides a listing of component names corresponding to each MLB-2 location.

ATTACHMENT 10 provides a listing of sequenced loads. 1 Check all the following MLB-2 1 Verify associated component indicating lights lit. status. 1 2 3 4 5 6 7 8 9 10 CTMTISO 3657 3198A 3772A 8112 LCV1003 7126 CONTRM CONTRM 3622 1 PHASE A CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED FILTFAN PRZN FAN CLOSED 1AON 1AON 2 3234A 3660 31980 37728 6149A 3377 3103 3104 3&49A 3624 CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED 3 Pl6V515 33188 2B66C 3772C 81496 3380 8033 3765 36496 3626 I CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED I 4 P16V517 3999A 2867C 3443 8149C 8871 8028 3766 3649C 3628 CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED 11 12 13 14 15 16 17 18 19 20 CTMTISO 3658 31988 3196 8100 7136 3331 CONTRM CONTRM 1 PHASE A CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED FILT FAN PRZNFAN 180N 1S0N i 2 32348 3198C 3197 8152 3376 3332 3623 3627 CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED f--_. 3 P16V514 33lSA 28660 3067 8880 7150 3333 3625 3629 CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED 4 P16V516 39998 2867D 3095 8860 8961 3334 8047 3659 CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSEO 2 Notify control room of phase A containment isolation status. -END-Page 1 of 5
12/11/2009 15 :5:; FNP-1-EEP-0 REACTOR TRIP OR SAFETY INJECTION Revision 37 LOCATION COMPONENT NUMBER 1-1 1-2 1-3 1-4 2-1 2-2 2-3 2 4 3-1 3-2 3-3 3-4 4-1 4-2 4-3 4-4 5-1 5-2 5-3 5-4 6-1 6-2 N/A Q1N12HV3234A QlP16V515 QIP16V517 QIE14HV3657 QIE14MOV3660 QIE14MOV3318B QIE12HV3999A QlP13HV3198A (QlPI3V284) QlP13HV3198D (QIPI3V281) QlP13HV2866C (Q1P13V301) QlP13HV2867C (QIP13V303) QIN25HV3772A Q1N25HV377 2B QIN25HV3772C QlP17HV3443 QIE21MOV8112 QIE21HV8149A Q1E21HV8149B QIE21HV8149C QIG21LCVI003 QIG21HV3377 ATTACHMENT 3 FIGURE 1 NAME CTMT ISO PHASE A TDAFWP STM SUPP WARMUP ISO (BOP) SW TO TURB BLDG ISO A TRN SW TO TURB BLDG ISO B TRN CTMT ATMOS TO R-11/12 ISO (BOP) CTMT ATMOS TO R-ll/12 ISO (BOP) CTMT ISO (BOP) RX CAV CLG DMPR (BOP) CTMT PURGE DMPRS (HS-3198) CTMT PURGE DMPRS (HS-3198) MINI-PURGE SUPPLY DAMPER MINI-PURGE EXHAUST DAMPER CHEM ADD TO lA SG ISO (BOP) CHEM ADD TO 1B SG ISO (BOP) CHEM ADD TO lC SG ISO (BOP) CCW FROM EXC LTDN/RCDT HXS RCP SEAL WTR RTN ISO LTDN ORIF ISO 45 GPM LTDN ORIF ISO 60 GPM LTDN ORIF ISO 60 GPM RCDT LeV CTMT SUMP DISCH (BOP) Page 2 of 5
1. RMS-40305A07 031 Page: 1 A LOCA has occurred on Unit 2. A review of the radiation monitor trends for RE-11 and RE-12 on the Integrated Plant Computer (I PC) shows an initial upscale followed by a slow trend towards background levels. Which ONE of the following describes the reason for the observed trend towards a background count rate? RE-11 and RE-12 are isolated from containment by a 'T'signal.
B. RE-11 and RE-12 are isolated from containment by a 'P' signal. C. RE-11 and RE-12 are isolated from containment by a'S' signal. D. RE-11 and RE-12 are isolated from containment by a Containment Ventilation Isolation signal. EEP-O ATTACHMENT 2 A. Correct, CTMT ATMOS TO R-11/12 01E14MOV3660,3657 &3658 are closed by a containment isolation -'T' signal. B. Incorrect, Containment phase B occurs at a higher pressure and does not affect the R-11/R-12 valves. C. Incorrect, CTMT ATMOS TO R-11/12 01 E14MOV3660,3657 &3658 close on a 'T' signal not an'S' signal. D. Incorrect, Containment ventilation isolation signal is generated by a MANUAL Phase A or B signal, any signal that generates an SI, HI-HI rad on RE-24A1B and does not affect the isolation valves for R-11/12. Rewrote the question and modified the distractors -tgb 4/30109 HL T-28 AUDIT EXAM 9/25/2009}}

ML101650244

Contents

Text

Reference:

References:

Reference:

References:

References:

References:

SUMMARY

References:

References:

References:

References:

References:

REFERENCES:

Reference:

References:

Reference:

2.0 INTRODUCTION

References:

References:

Reference:

REFERENCES:

References:

References:

References:

References:

References:

References:

Navigation menu

ML101650244

Text

Reference:

References:

Reference:

References:

References:

References:

SUMMARY

References:

References:

References:

References:

References:

REFERENCES:

Reference:

References:

Reference:

2.0 INTRODUCTION

References:

References:

Reference:

REFERENCES:

References:

References:

References:

References:

References:

References:

Navigation menu

Search